WorldWideScience

Sample records for vver reactor pressure

  1. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  2. 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.

  3. 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)

  4. 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.

  5. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  6. Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Nikonov, S. [NRC ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2014-08-15

    The paper presents an uncertainty and sensitivity (U and S) study of the VVER-1000 reactor hydraulic properties. It is based on the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). The novelty of the work consists of taking into consideration all hydraulic uncertainty parameters used in the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel ATHLET model of the RPV is developed. It consists of ca. 26 600 control volumes most of them connected with junctions for cross flows. The specific geometry of the gap between upper part of the baffle and upper part of fuel assembly and also a fuel assembly head is taken explicitly into account The influence of the input parameters on the sensitivity and uncertainty of the RPV outlet and inlet temperatures and mass flows as well assembly-wise mass flow and coolant temperature axial distributions is shown.

  7. Development and validation of coupled dynamics code 'TRIKIN' for VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Obaidurrahman, K; Doshi, J. B.; Jain, R. P. [IIT Bombay, Mumbai (India); Jagannathan, V. [Bhabha Atomic Research Centre, Mumbai (India)

    2010-06-15

    New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors

  8. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: laurahg42@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2014-10-15

    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  9. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  10. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  11. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    Science.gov (United States)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  12. VVER Reactor Safety in Eastern Europe and Former Soviet Union

    Science.gov (United States)

    Papadopoulou, Demetra

    2012-02-01

    VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

  13. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  14. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  15. Influence of operation factors on brittle fracture initiation and critical local normal stress in SE(B) type specimens of VVER reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Erak, A. D.; Kiselev, A. S.; Bubyakin, S. A.; Bandura, A. P.

    2015-12-01

    A complex of mechanical tests and fractographic studies of VVER-1000 RPV SE(B) type surveillance specimens was carried out: the brittle fracture origins were revealed (non-metallic inclusions and structural boundaries) and the correlation between fracture toughness parameters (CTOD) and fracture surface parameters (CID) was established. A computational and experimental method of the critical local normal stress determination for different origin types was developed. The values of the critical local normal stress for the structural boundary origin type both for base and weld metal after thermal exposure and neutron irradiation are lower than that for initial state due to the lower cohesive strength of grain boundaries as a result of phosphorus segregation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  17. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  18. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Science.gov (United States)

    Koš'ál, Michal; Rypar, Vojtìch; Harutyunyan, Davit; Schulc, Martin; Losa, Evžen

    2017-09-01

    The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  19. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  20. Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

    Directory of Open Access Journals (Sweden)

    О. M. Pugach

    2015-07-01

    Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

  1. Pressure Vessel Calculations for VVER-440 Reactors

    Science.gov (United States)

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

    2003-06-01

    Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

  2. EVALUATION METRICS APPLIED TO ACCIDENT TOLERANT FUEL CLADDING CONCEPTS FOR VVER REACTORS

    Directory of Open Access Journals (Sweden)

    Martin Sevecek

    2016-12-01

    Full Text Available Enhancing the accident tolerance of LWRs became a topic of high interest in many countries after the accidents at Fukushima-Daiichi. Fuel systems that can tolerate a severe accident for a longer time period are referred as Accident Tolerant Fuels (ATF. Development of a new ATF fuel system requires evaluation, characterization and prioritization since many concepts have been investigated during the first development phase. For that reason, evaluation metrics have to be defined, constraints and attributes of each ATF concept have to be studied and finally rating of concepts presented. This paper summarizes evaluation metrics for ATF cladding with a focus on VVER reactor types. Fundamental attributes and evaluation baseline was defined together with illustrative scenarios of severe accidents for modeling purposes and differences between PWR design and VVER design.

  3. Steam Line Break investigation at full power reactor for VVER-1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Pavlova, M., E-mail: pavlovamp@mail.bg; Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg

    2015-04-15

    Highlights: • In this study we investigated Steam Line Break accident at full power reactor. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP/MOD 3.2 computer code is used in performing the analyses. • The results are used for analytical validation of EOP. - Abstract: This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. The RELAP5/MOD 3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia. The main purpose of the analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant's state and the critical safety function (CSF). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The performed analysis is based on a previously used bounding approach in analytical validation of SB EOPs. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The presented thermal-hydraulic calculations of the accident scenarios involve the loss of CSF “Subcriticality” for VVER-1000/V320 units at KNPP.

  4. The influence of operational and water chemistry parameters on the deposits of corrosion products on fuel assemblies at nuclear power plants with VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Rodionov, Yu. A.; Gavrilov, A. V.

    2011-07-01

    The phenomenon involving a growth of pressure drop in the reactor core and redistribution of deposits in the reactor core and primary coolant circuit of a nuclear power station equipped with VVER-440 reactors is considered. A model is developed, the physicochemical foundation of which is based on the dependence of corrosion product transfer on the temperature and pH t value of coolant and on the correlation between the formation rate of corrosion products (Fe) (after subjecting the steam generators to decontamination) and rate with which they are removed from the circuit. The purpose of the simulation carried on the model is to predict the growth of pressure drop on the basis of field data obtained from nuclear power installations and correct the water chemistry (by adjusting the concentrations of KOH, H2, and NH3) so as to keep the pressure drop in the reactor at a stable level.

  5. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    Directory of Open Access Journals (Sweden)

    Tikhomirov Georgy

    2017-01-01

    Full Text Available For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

  6. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  7. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  8. Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation: material: 15 Ch2MFA {approx} 0, 15 C-2,5 Cr-0, 7Mo-0,3 V; Veraenderungen der Mikrostruktur in Grundwerkstoff, WEZ und Schweissgut eines 440-VVER-Reaktordruckbehaelters, verursacht durch Neutronenbestrahlung im langzeitigen Betrieb; Werkstoff: 15 Ch2MFA {approx} 0,15 C-2,5 Cr-0, 7Mo-0,3 V

    Energy Technology Data Exchange (ETDEWEB)

    Maussner, G.; Scharf, L.; Langer, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Gurovich, B. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1998-11-01

    Within the scope of the Tacis `91/1.1 project of the European Community, ``Reactor Vessel Embrittlement``, specimens were taken from the heavily irradiated circumferential welds of a VVER pressure vessel. The cumulated fast neutron fluence in the specimens amounts to up to 6.5 x 10{sup 19} cm{sup -}2 (E > 0.5 MeV). For the multi-laboratory, coordinated study, the specimens were cutted for mechanical testing as well as analytical, microstructural and microanalytical examinations in the base metal, HAZ and weld metal with respect to the effects of reactor operatio and post-irradiation annealing as well as thermal treatment (475 C, 560 C). The analytical transmission electron microscopy (200 kV) revealed the alterations found in the mechanical properties to be due to the formation of black dots and irradiation-induced segregations and accumulations of copper and carbides. These effects, caused by operation, (neutron radiation, temperature), are much more significant in the HAZ than in the base metal. (orig./CB) [Deutsch] Im Rahmen des von der Europaeischen Union beauftragten Tacis `91/1.1 Programms `Reactor Vessel Embrittlement` wurden Bohrkerne aus dem hochbestrahlten Rundnahtbereich eines VVER-Reaktordruckbehaelters entnommen. Die kumulierte schnelle Neutronenfluenz in diesen Proben betraegt bis zu 6,5 x 10{sup 19} cm{sup -2} (E>0,5 MeV). In einer gemeinschaftlichen Untersuchung wurden mechanisch-technologische, chemische sowie mirkostrukturelle Untersuchungen an Grundwerkstoff-, WEZ- und Schweissgutproben im vergleichbaren Ausgangs-, bestrahlten und anschliessend waermebehandelten (475 C, 560 C) Werkstoffzustand durchgefuehrt. Die analytische Durchstrahlelektronenmikroskopie (200 kV) laesst als Ursache fuer die festgestellten Veraenderungen der mechanischen Eigenschaften die Bildung von Versetzungsringen (black dots) sowie von bestrahlungsinduzierten Ausscheidungen und Anreicherungen von Kupfer in den Karbiden erkennen. Diese Effekte, als Folge der betrieblichen

  9. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  10. Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

    2000-03-01

    The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

  11. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  12. Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

    Science.gov (United States)

    Hölgye, Z; Filgas, R

    2006-04-01

    Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

  13. Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

    Science.gov (United States)

    Tyapkov, V. F.; Erpyleva, S. F.

    2017-05-01

    Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

  14. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  15. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  16. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  17. Estimation of material degradation of VVER-1000 baffle

    Science.gov (United States)

    Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

    2017-09-01

    The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  18. Estimation of material degradation of VVER-1000 baffle

    Directory of Open Access Journals (Sweden)

    Harutyunyan Davit

    2017-01-01

    Full Text Available The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  19. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  20. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  1. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  2. Simulation of a control rod ejection in a VVER-1000 reactor with the program Relap5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.; Gehin, J.C.; Yoder, G.L. [Oak Ridge National Lab., TN (United States); Ivanov, V.K. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    2001-07-01

    The RELAP5-3D code has been employed to simulate the ejection of a control rod at the Balakovo-4 plant, a VVER-1000 V320 plant located in Russia. The reactor core contains 163 assemblies, three of them Lead Test Assemblies (LTAs) with mixed oxide (MOX) fuel, and the remaining 160 assemblies with UO{sub 2} fuel. The worth of the ejected control rod was $ 0,225 or 142 pcm. Results from point and three-dimensional (3-D) or nodal kinetics calculations are presented. The results from both models are similar with no significant differences. All calculated results are within safety limits, with no fuel melting or cladding failures predicted to occur. (author)

  3. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  4. The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

    Energy Technology Data Exchange (ETDEWEB)

    Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov [State Scientific Center of Russian Federation - Institute for Physics and Power Engineering by A.I. Leypunsky, IPPE, 1, Bondarenko sq., Obninsk, 249030 (Russian Federation); Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko [FSUE Atomenergoproekt, B-5, 1 bldg.7, Bakuninskaya, 107005, Moscow (Russian Federation); Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev [FSUE OKB Gidropress, 21, Ordzhonikidze street, 142103 Podolsk (Russian Federation)

    2005-07-01

    Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m{sup 3}. The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

  5. Prospects of VVER-SKD reactor in a closed fuel cycle

    OpenAIRE

    Glebov, A.P.; Klushin, A.V.; Yu.D. Baranaev

    2015-01-01

    At the new centure's begin eight countries with developed nuclear power industry took part under the aegis of the IAEA in research of innovative nuclear reactors and fuel cycles to choose a nuclear power system with fast reactors based on a closed fuel cycle (CFC) and to perform joint R&D in this direction. An agreement was reached on the use of based on proven technologies CNFC-FR (Closed Nuclear Fuel Cycles and Fast Reactors), as a reference system for common assessment. Common principle...

  6. Development and application of the coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2006-07-01

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)

  7. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    OpenAIRE

    Ma, Weimin; Yuan, Yidan; Sehgal, Bal Raj

    2016-01-01

    A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential...

  8. 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.)

  9. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors

    1998-03-01

    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  10. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  11. VVER Knowledge Preservation and Transfer within the Frame of CORONA Project Activities

    Science.gov (United States)

    Mitev, Mladen; Corniani, Enrico; Manolova, Maria; Pironkov, Lybomir; Tsvetkov, Iskren

    2016-02-01

    The CORONA project is funded by the European Commission under the FP7 programme with overall objective to establish a Regional Centre of Competence for VVER Technology and Nuclear Applications. The Centre will provide support and services for preservation and transfer of VVER-related nuclear knowledge as well as know-how and capacity building. Specific training schemes aimed at nuclear professionals and researchers, non-nuclear professionals and students are developed and implemented in cooperation with local, national and international training and educational institutions. Pilot trainings are executed for each specific target group to assess the applicability of the training materials. The training scheme implemented for nuclear professionals and researchers is focussed on the NPP Lifetime Management. The available knowledge on enhancing safety and performance of nuclear installations with VVER technology is used in the preparation of the training materials. The Online Multimedia Training Course on VVER Reactor Pressure Vessel Embrittlement and Integrity Assessment, developed by the joint effort of JRC-IET and IAEA is used in the training. The outcome collected from the trainees showed that the tool meets its primary goal of consolidating the existing knowledge on the VVER RPV Embrittlement and Integrity Assessment, provides adequate ground for transfer of this knowledge.

  12. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)], Sokolov, M.A., Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

    1997-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVS) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The working group agreed that each side would irradiate, anneal, reirradiate (if feasible), and test two materials of the other; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) with two weld metals representative of VVER-440 AND VVER-1000 RPVS, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation,annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) program plate 02 and Heavy-Section Steel Irradiation (HSSI) program weld 73w. The results for each material from each laboratory are compared with those from the other laboratory. the ORNL experiments with the VVER welds included irradiation to about 1 x 10 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  13. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  14. VVER-1000 MOX Core Computational Benchmark: Specification and Results

    National Research Council Canada - National Science Library

    Mikhail Kalugin; Eugeny Gomin; Dmitry Oleynik

    2006-01-01

    This report presents the VVER MOX Core Computational Benchmark Specification and Results, which was proposed as a benchmark within the OECD/NEA Expert Group on Reactor-based Plutonium Disposition (TFRPD...

  15. Use of gadolinium burnable absorbers in VVER Type Reactors. Validation of WIMS-D/4 code; Empleo del gadolinio como absorbente quemable en los reactores nucleares VVER. Validacion del codigo WIMS-D/4

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez Cardona, Caridad M.; Guerra Valdes, Ramiro; Lopez Aldama, Daniel [Centro de Tecnologia Nuclear, La Habana (Cuba)

    1996-07-01

    Burnable absorbers are not used in current operating WWERs, but in order to optimize the fuel cycle and enhance operational safety, one should also introduce gadolinium or a similar burnable absorber in these reactors. For this purpose adequate tools for properly calculating local effects in hexagonal geometries should be developed and validated. The present gives main results in validating the WIMS-D/4 lattice code for Gd burnable absorber bearing WWER lattices. To validate the code experimental and calculational benchmarks proposed in a IAEA Coordinated Research Program were solved. A code system for the optimization of the Gd axial distribution in a WWER reactor was developed and it also presented here. (author)

  16. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  17. Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

    Science.gov (United States)

    Parchevsky, V. M.; Guryanova, V. V.

    2017-01-01

    A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

  18. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    CSIR Research Space (South Africa)

    Hirschberg, G

    1999-03-01

    Full Text Available Ag deposition is needed, which may facilitate the elaboration of more e cient surface prevention and/or decontamination procedures. As demonstrated above, corrosion and contamina- tion processes in the primary cooling circuit of PWRs are essentially interrelated...: the contaminant isotopes are mostly corrosion products activated in the reactor core, and the contamination takes place on surfaces which were modi?ed by the corrosion. Also, the two counter measures (decontamination and corrosion-prevention) are connected to each...

  19. Reactor pressure vessel. Status report

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  20. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-06-16

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  1. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  2. Validation of finite difference core diffusion calculation methods with FEM and NEM for VVER-1000 MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); RPDD, Central Complex, BARC, Mumbai - 400085 (India); Singh, T. [Reactor Physics and Nuclear Engineering Section, Reactor Group, BARC, Mumbai (India); Pal, U.; Karthikeyan, R. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); Sundaram, G. [Nuclear Safety Group, KK-NPC, Mumbai (India)

    2006-07-01

    India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)

  3. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    Directory of Open Access Journals (Sweden)

    Weimin Ma

    2016-03-01

    Full Text Available A historical review of in-vessel melt retention (IVR is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs. The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

  4. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Maltsev, D. A.; Fedotova, S. V.; Frolov, A. S.; Zhuchkov, G. M.

    2017-01-01

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

  5. Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

  6. CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model

    Directory of Open Access Journals (Sweden)

    F. Moretti

    2009-01-01

    Full Text Available A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of EC TACIS project R2.02/02: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration at core inlet.” Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle, and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for nuclear reactor safety. Both a pretest and a posttest CFD simulations of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, resp.. The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

  7. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  8. Construction management of Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bohra, S.A. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)]. E-mail: sabohra@npcil.co.in; Sharma, P.D. [Nuclear Power Corporation of India Limited, Vikram Sarabhai Bhavan, Anushaktinagar, Mumbai 400094 (India)

    2006-04-15

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  9. Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Moretti, F.; Melideo, D. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Del Nevo, A., E-mail: delnevo@hotmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); D' Auria, F. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Hoehne, T. [Forschungszentrum Dresden-Rossendorf (FZD), P.O.B. 51 01 19, D-01314 Dresden (Germany); Lisenkov, E. [FSUE OKB Gidropress, Ordshonikidize 21, RU-142103 Podolsk, Moscow district (Russian Federation); Gallori, D. [AREVA NP SAS, Tour AREVA - 92084 Paris, La Defense Cedex (France)

    2011-08-15

    Highlights: > Five mixing experiments in a scaled model of a VVER-1000 are described and discussed. > In-vessel mixing investigations of the coolant properties distribution at the core inlet. > These tests brought an improvement to existing experimental database for TH code validation. - Abstract: In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project 'TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet'. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.

  10. ANDREA 2.2 and 2.3. Advances in modelling of VVER cores

    Energy Technology Data Exchange (ETDEWEB)

    Havluj, Frantisek; Hejzlar, Jonatan; Vocka, Radim; Vysoudil, Jiri [UJV Rez, Husinec-Rez (Czech Republic)

    2017-09-15

    In 2016 a new version of code ANDREA for core design and reload safety analysis of VVER reactors has been released. The new code version includes several major improvements. The first of them is a seamless incorporation of short time kinetics calculations (without temperature feedback) into the code. This new feature accompanied by the possibility of excore detector signal predictions enables precise interpretation of dynamic measurements of control assembly weight during the reactor startup. Second important enhancement resides in new flexible format of cross section libraries and in new fuel temperature model based on results of TRANSURANUS fuel performance code. The new code version has been thoroughly tested and validated for both VVER440 and VVER-1000 reactors. Furthermore for the new version 2.3 which is to be released shortly we have implemented the possibility of fluent control assemblies' motion and of non-equidistant axial nodalization schemes in VVER-440 calculations.

  11. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-01

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

  12. 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.

  13. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  14. Analytical validation of operator actions based on SAMG for VVER 1000 with ASTECv2r3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-01-15

    Highlights: • Performing of analytical validation of operator action based SAMG. • Simulation of base calculation of SBO scenario without operator action for VVER 1000. • Simulation of SBO scenario with investigation of operator actions based on SAMG for VVER 1000. - Abstract: This paper presents the analytical validation of operator action based on severe accident management guidelines (SAMG) for Kozloduy NPP VVER1000 with severe accident computer code ASTECv2r3. The work is oriented on investigation of plant behavior during total loss of power and the operator actions performed based on strategies considered in severe accident management guidelines (SAMG) in Kozloduy nuclear power plant (KNPP). Using the SAMG strategies the operator depressurize primary circuit by gas removing system (YR) and try to cool down the reactor core by high pressure injection system (HPIS). The purpose of these analyses is to examine the possibility of keeping the core from further damage during a severe accident and to assess the likelihood of additional generation of hydrogen by additional flooding of the heated core. For this purpose it have been simulated a SBO scenario with injection of cold water by a high pressure pump (HPP) in cold leg at different core exit temperatures at 923 K and 1253 K. The selection of investigated analyses was based on severe accident management strategy of KNPP VVER1000. The presented work is important for analytical validation, verification, and further improvements of SAMG as well as for assessment of Level 2 probabilistic safety analyses (L2 PSA). The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research.

  15. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    Energy Technology Data Exchange (ETDEWEB)

    Foehl, J.; Weissenberg, T. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) (Spain); Ernestova, M.; Zamboch, M. [Nuclear Research Inst. (NRI) (Czech Republic); Seifert, H.P.; Ritter, S. [Paul Scherrer Inst. (PSI) (Switzerland); Roth, A.; Devrient, B. [Framatome ANP GmbH (F ANP) (Germany); Ehrnsten, U. [Technical Research Centre of Finland (VTT) (Finland)

    2004-07-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  16. IMPROVED MODELS AND METHOD OF POWER CHANGE OF NPP UNIT WITH VVER-1000

    Directory of Open Access Journals (Sweden)

    Tymur Foshch

    2017-05-01

    Full Text Available This study represents the improved mathematical and imitational allocated in space multi-zone model of VVER-1000 which differs from the known one. It allows to take into account the energy release of 235U nuclei fission as well as 239Pu . Moreover, this model includes sub-models of simultaneous control impact of the boric acid concentration in the coolant of the first circuit and the position of 9th group control rods which allows to consider it as the model with allocated parameters and also allows to monitor changes in the mentioned technological parameters by reactor core symmetry sectors, by layers of reactor core height and by fuel assembly group each symmetry sector. Moreover, this model allows to calculate important process-dependent parameters of the reactor (including axial offset as quantitative measure of its safety. As the mathematical and imitational models were improved, it allows to take into account intrinsic properties of the reactor core (including transient processes of xenon and thus reduce the error of modelling static and dynamic properties of the reactor.The automated control method of power change of the NPP unit with VVER-1000 was proposed for the first time. It uses three control loops. One of which maintains the regulatory change of reactor power by regulating the concentration of boric acid in the coolant, the second circuit keeps the required value of axial offset by changing the position of control rods, and the third one holds constant the coolant temperature mode by regulating the position of the main turbo generator valves.On the basis of the above obtained method, two control programs were improved. The first one is the improved control program that implements the constant temperature of the coolant in the first circuit and the second one is the improved control program that implements the constant steam pressure in the second circuit.

  17. The vver severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V. [Russian Research Center, Kurchatov Institute, Moscow (Russian Federation)

    2001-07-01

    The basic approach to the VVER safety management is based on the defence-in-depth principle the main idea of which is the multiplicity of physical barriers on the way of dangerous propagation on the one hand and the diversity of measures to protect each of them on the other hand. The main events of severe accident with loss of core cooling at NPP with WWER can be represented as a sequence of NPP states, in which each subsequent state is more severe than the previous one. The following sequence of states of the accident progression is supposed to be realistic and the most probable: -) loss of efficient core cooling; -) core melting, relocation of the molten core to the lower head and molten pool formation, -) reactor vessel damage, and -) containment damage and fission products release. The objectives of accident management at the design basis stage, the determining factors and appropriate determining parameters of processes are formulated in this paper. The same approach is used for the estimation of processes parameters at beyond design basis accident progression. The accident management goals and the determining factors and parameters are also listed in that case which is characterized by the loss of integrity of the fuel cladding. The accident management goal at the stage of core melt relocation implies the need for an efficient core-catcher.

  18. Nuclear reactor pressure vessel support system

    Science.gov (United States)

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  19. 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.

  20. Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors - A case study

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Usha [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: ushapal@barc.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: vjagan@barc.gov.in

    2008-10-15

    Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.

  1. Technology of repair of selected equipment in the power plant type VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Barborka, J.; Magula, V. [Welding Research Inst. (WRI), Bratislava (Slovakia)

    1998-11-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored.

  2. Methodological studies on the VVER-440 control assembly calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hordosy, G.; Kereszturi, A.; Maraczy, C. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

    1995-12-31

    The control assembly regions of VVER-440 reactors are represented by 2-group albedo matrices in the global calculations of the KARATE code system. Some methodological aspects of calculating albedo matrices with the COLA transport code are presented. Illustrations are given how these matrices depend on the relevant parameters describing the boron steel and steel regions of the control assemblies. The calculation of the response matrix for a node consisting of two parts filled with different materials is discussed.

  3. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  4. Cascading pressure reactor and method for solar-thermochemical reactions

    Energy Technology Data Exchange (ETDEWEB)

    Ermanoski, Ivan

    2017-11-14

    Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

  5. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  6. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  7. Undermoderated spectrum MOX core study. Supercritical pressure light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki; Kosizuka, Seiichi [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    1998-09-01

    The supercritical pressure light water cooling reactor is a nuclear reactor concept with the once through type and the direct cycle reactor cooled with supercritical pressure water. The cooling water controlled with the feed pump flows directly to the turbine and a recirculation is never done by the nuclear reactor of this type. Therefore, this system isn`t equipped with the recirculation system and the steam separator, the system becomes simple. As for this system, it is expected that the cost performance improves. Here, the outline of former study is described. (author)

  8. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  9. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  10. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  11. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  12. Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project

    Science.gov (United States)

    Asmolov, V. G.; Zagryazkin, V. N.; Tsurikov, D. F.; Vishnevsky, V. Yu.; D'Yakov, Ye. K.; Kotov, A. Yu.; Repnikov, V. M.

    2010-12-01

    The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.

  13. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  14. Study of non-condensable gases effect on VVER steam generator operation in condensation mode at large-scale facility

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, A. V.; Remizov, O. V.; Tzyganok, A. A.; Kalyakin, D. S., E-mail: sas@ippe.r [Institute for Physics and Power Engineering by A. I. Leypunsky, Bondarenko 1 sq. Obninsk, 249033 Moscow (Russian Federation)

    2010-10-15

    The NNP-2006 project of nuclear power plant with VVER-1200 reactor provides for use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (Sg) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from Sg arrives at the core providing its additional cooling. The joint operation of the PHRS and the system of hydro accumulators of the second stage makes it possible to assure the heat removal from the core during 24 hours. The efficiency of the system can be affected by the presence os non-condensable gases in the primary circuit. The main sources of gases are nitrogen, arriving at the circuit, as hydro accumulators actuate and products of radiolysis of water. The circuit design considered in the project makes it possible to remove gas-steam mixture from Sg. At the same time, it is necessary to ascertain if the gas removal is adequate for ensuring the design operation of Sg in the steam condensation mode. For this purpose, series of experiments have been carried out at the large-scale test facility HA2M-Sg. The test facility incorporates VVER reactor Sg model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator and buffer tank, equipped by steam supply system. The elevations of the main equipment correspond to those of reactor project. Experiments at the HA2M-Sg test facility have been performed at the pressure 0.36-0.38 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the basic results of experiments aimed at the evaluation of Sg condensation power under the inflow of gas-steam mix to the tube bundle, both under the simulation of gas-steam mixture outflow from Sg and without outflow. (Author)

  15. The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

    Energy Technology Data Exchange (ETDEWEB)

    Kot, C.

    1999-05-10

    Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.

  16. 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.

  17. Pressurized hydrogenotrophic denitrification reactor for small water systems.

    Science.gov (United States)

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal

    2017-03-15

    The implementation of hydrogenotrophic denitrification is limited due to safety concerns, poor H2 utilization and low solubility of H2 gas with the resulting low transfer rate. The current paper presents the main research work conducted on a pressurized hydrogenotrophic reactor for denitrification that was recently developed. The reactor is based on a new concept suggesting that a gas-liquid equilibrium is achieved in the closed headspace of denitrifying reactor, further produced N2 gas is carried out by the effluent and gas purging is not required. The feasibility of the proposed reactor was shown for two effluent concentrations of 10 and 1 mg NO3--N/L. Hydrogen gas utilization efficiencies of 92.8% and 96.9% were measured for the two effluent concentrations, respectively. Reactor modeling predicted high denitrification rates above 4 g NO3--N/(Lreactor·d) at reasonable operational conditions. Hydrogen utilization efficiency was improved up to almost 100% by combining the pressurized reactor with a following open-to-atmosphere polishing unit. Also, the potential of the reactor to remove ClO4- was shown. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. RISKAUDIT Report no. 7, Vol. 1: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

  19. 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)

  20. Design of virtual SCADA simulation system for pressurized water reactor

    Science.gov (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  1. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  3. Corrosion product behavior in VVER secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V.A.; Velikopolsky, S.V.; Yurmanov, E.V. [N.A. Dollezhal Research and Development Inst. of Power Engineering (NIKIET), Moscow (Russian Federation)

    2010-07-01

    Accumulation of corrosion products lead to some problems during long-term operation of VVER plants, such as secondary system component degradation including crud-induced local corrosion and corrosion cracking. Corrosion sludge and deposit removal from steam generators and other equipment is costly and time-consuming and leads to additional waste production. This problem is vital in the case of plant life extension. Appropriate solutions of the problem could be developed based on both Russian and international experience of the VVER fleet. Recommendations on how to mitigate corrosion product accumulation in VVER secondary systems were developed based on comparative analysis of available long-term data on corrosion product behavior in all the operating VVER plants, such as the following: Sludge and deposit accumulation in inner surfaces of secondary piping and components; Corrosion rate measurements using in-situ specimen testing at operated VVER plants; Efficiency of corrosion product removal from secondary system water by means of condensate polishers and steam generator blowdown cleanup systems; Sludge and deposit removal from steam generators during chemical cleaning; Secondary piping and components conservation efficiency during long outages. Comparative data analysis of corrosion product behavior has shown different corrosion product accumulation rates in Novovoronezh, Kola, Kalinin, Balakovo and Rostov NPPs. The said difference is due to different design and operation peculiarities. (author)

  4. Reactor pressure vessel head vents and methods of using the same

    Science.gov (United States)

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  5. 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.

  6. Dominant accident sequences in Oconee-1 pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dearing, J F; Henninger, R J; Nassersharif, B

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling.

  7. Diagnostics and control of pressurized reactors using artificial neural networks

    Science.gov (United States)

    Ikonomopoulos, Andreas; Tsoukalas, Lefteri H.; Uhrig, Robert E.

    1992-09-01

    A methodology employing artificial neural networks and fuzzy arithmetic in the diagnosis and control of complex systems such as pressurized water reactors is presented. Fuzzy numbers represent the linguistic values of plant-specific variables, e.g., performance or availability. The notion of a virtual instrument, i.e., a software-based measuring device calibrated to the idiosyncrasies of a specific system is used. Neural networks perform a mapping of physically measurable parameters to fuzzy numbers called Virtual Measurement Values (VMV). The methodology is tested with start-up data from an experimental nuclear reactor. The results demonstrate the very good capacity of such virtual instruments for failure-tolerance and suggest the possibility of developing alternative algorithms for diagnostics and control.

  8. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  9. Analysis of loss of off-site power ATWS in VVER-440 concept

    Energy Technology Data Exchange (ETDEWEB)

    Hoeppner, G.; Siltanen, P.; Kotro, J.

    1987-01-01

    During 1985 the Finnish state-owned utility Imatran Voima Oy signed a work order with Gesellschaft fuer Reaktorsicherheit mbH of the Federal Republic of Germany (GRS) for the analysis of abnormal transients in a pressurized water reactor (PWR) concept based on a Soviet design. The results of these calculations were intended to be introduced into the licensing process and to support a decision to build such a nuclear power station. A computer model was constructed of the VVER-440 concept, a 500-MW(electric) PWR designed in the USSR and modified for Finland. The ALMOD4 code, developed at GRS, was used for the investigation. The ALMOD4 code is a fast running code for the analysis of operational and abnormal transients in PWRs. Input data were set up to calculate anticipated transients without scram, most notably the loss of off-site power case. One-dimensional neutron kinetics was used to correctly model the neutronics feedback of axially distributed moderator density and fuel temperature in a changing axial power profile. Interlocking signals and the engineered safety systems were modeled to assess the overall systems response to this abnormal transient. Special analytical problems were encountered since a detailed and verified model of the steam generator (SG) with horizontally positioned heat exchanger tubes was not available. Therefore, two bounding calculations were performed with different SG models.

  10. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  11. Advanced fuels for plutonium management in pressurized water reactors

    Science.gov (United States)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  12. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-15

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

  13. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  14. VVER-1000 MOX Core Computational Benchmark analysis using indigenous codes EXCEL, TRIHEX-FA and HEXPIN

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Sunil sunny, C.; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-10-15

    Validation studies based on the analysis of theoretical benchmarks play a key role in the identification of deficiencies in the reactor physics design computational codes and the associated nuclear data libraries. Implementation of improvements, if any, in theoretical models and the choice of appropriate nuclear data libraries help in enhancing the accuracy of calculations. As part of the effort for the validation of computer codes for plutonium utilization in VVER type reactors, the indigenous codes EXCEL, TRIHEX-FA and HEXPIN, developed at Light Water Reactor Physics Section (LWRPS), RPDD, BARC, and the associated nuclear data library (JEF22XS), were employed to analyse 'VVER-1000 MOX Core Computational Benchmark'. The few group homogenized parameters of assembly cell or individual lattice cells were obtained by the hexagonal lattice burn-up code EXCEL and the core diffusion calculations were then performed using hexagonal assembly geometric code TRIHEX-FA or the pin-by-pin diffusion code HEXPIN. VVER-1000 reactor core loaded with 2/3rd of Low-Enriched Uranium (LEU) fuel assemblies (FAs) and 1/3rd of weapons grade MOX FAs was investigated. Effective multiplication factors and assembly average fission reaction rate distributions have been calculated for various reactor state descriptions using 3-D diffusion theory codes TRIHEX-FA and HEXPIN. Further, estimate of detailed pin-by-pin fission reaction rate distributions of a few selected assemblies were made for the normal working state of the reactor using pin-by-pin core simulation code HEXPIN. A comparison of results was done with the reported Monte Carlo (MC) values of the benchmark and in most cases good agreement was observed with the benchmark results.

  15. Advanced power plant training simulator for VVER-440/V230 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shier, W.; Kennett, R. [Brookhaven National Lab., Upton, NY (United States); Vaclav, E.; Gieci, A. [Nuclear Power Research Inst. Trnava, Inc. (Slovakia)

    1996-11-01

    An advanced, workstation based, nuclear power plant simulator has been developed for use in training the operational staff of the Bohunice Nuclear Power Plant. This training simulator uses state-of- the-art computer hardware and software and provides the capability to simultaneously include six members of the power plant operating staff in the training sessions. A detailed reactor model has been developed, representing the Bohunice VVER-44O/V230 plants, for use with the RELAP5 simulation software. In addition, a comprehensive validation program has been completed that compares the simulation results of the advanced simulator with the results from a current VVER-44O/V230 simulator. A summary of the training features and capabilities of the simulator is also provided.

  16. Iron Catalyst Chemistry in High Pressure Carbon Monoxide Nanotube Reactor

    Science.gov (United States)

    Scott, Carl D.; Povitsky, Alexander; Dateo, Christopher; Gokcen, Tahir; Smalley, Richard E.

    2001-01-01

    The high-pressure carbon monoxide (HiPco) technique for producing single wall carbon nanotubes (SWNT) is analyzed using a chemical reaction model coupled with properties calculated along streamlines. Streamline properties for mixing jets are calculated by the FLUENT code using the k-e turbulent model for pure carbon monixide. The HiPco process introduces cold iron pentacarbonyl diluted in CO, or alternatively nitrogen, at high pressure, ca. 30 atmospheres into a conical mixing zone. Hot CO is also introduced via three jets at angles with respect to the axis of the reactor. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Cluster reaction rates are from Krestinin, et aI., based on shock tube measurements. Another model is from classical cluster theory given by Girshick's team. The calculations are performed on streamlines that assume that a cold mixture of Fe(CO)5 in CO is introduced along the reactor axis. Then iron forms clusters that catalyze the formation of SWNTs from the Boudouard reaction on Fe-containing clusters by reaction with CO. To simulate the chemical process along streamlines that were calculated by the fluid dynamics code FLUENT, a time history of temperature and dilution are determined along streamlines. Alternative catalyst injection schemes are also evaluated.

  17. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  18. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  19. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  20. 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.

  1. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  2. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  3. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  4. RISKAUDIT Report no. 7, Vol. 2: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The second part of the document covers the following aspects of the report: accident analysis; systems analysis; plant operation; operating experience feedback; radio protection and health; probabilistic safety assessment; summary and future plans.

  5. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  6. 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.

  7. Development of a cross-section methodology and a real-time core model for VVER-1000 simulator application

    Energy Technology Data Exchange (ETDEWEB)

    Georgieva, Emiliya Lyudmilova

    2016-06-06

    The novel academic contributions are summarized as follows. A) A cross-section modelling methodology and a cycle-specific cross-section update procedure are developed to meet fidelity requirements applicable to a cycle-specific reactor core simulation, as well as particular customer needs and practices supporting VVER-1000 operation and safety. B) A real-time version of the Nodal Expansion Method code is developed and implemented into Kozloduy 6 full-scope replica control room simulator.

  8. SEARCH FOR THE BEST POWER CONTROL PROGRAM AT NPP WITH VVER-1000 USING GRADIENT DESCENT METHOD

    Directory of Open Access Journals (Sweden)

    S. N. Pelykh

    2016-09-01

    Full Text Available This article is regarded to the search for the best power control program at nuclear power plant (NPP with VVER- 1000 by gradient descent method for the objective function, which includes the criteria of efficiency, safety and damage. Criteria normalization to the maximum value is carried out when looking for the minimum of the objective function because criteria have different physical nature. There were chosen such objective criteria as depth of fuel burn-up, index of the fuel cladding damage and axial offset - the ratio of the energy at the top and bottom of the reactor core.

  9. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    Science.gov (United States)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  10. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  11. 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 Criteria...

  12. High-performance simulations for atmospheric pressure plasma reactor

    Science.gov (United States)

    Chugunov, Svyatoslav

    Plasma-assisted processing and deposition of materials is an important component of modern industrial applications, with plasma reactors sharing 30% to 40% of manufacturing steps in microelectronics production. Development of new flexible electronics increases demands for efficient high-throughput deposition methods and roll-to-roll processing of materials. The current work represents an attempt of practical design and numerical modeling of a plasma enhanced chemical vapor deposition system. The system utilizes plasma at standard pressure and temperature to activate a chemical precursor for protective coatings. A specially designed linear plasma head, that consists of two parallel plates with electrodes placed in the parallel arrangement, is used to resolve clogging issues of currently available commercial plasma heads, as well as to increase the flow-rate of the processed chemicals and to enhance the uniformity of the deposition. A test system is build and discussed in this work. In order to improve operating conditions of the setup and quality of the deposited material, we perform numerical modeling of the plasma system. The theoretical and numerical models presented in this work comprehensively describe plasma generation, recombination, and advection in a channel of arbitrary geometry. Number density of plasma species, their energy content, electric field, and rate parameters are accurately calculated and analyzed in this work. Some interesting engineering outcomes are discussed with a connection to the proposed setup. The numerical model is implemented with the help of high-performance parallel technique and evaluated at a cluster for parallel calculations. A typical performance increase, calculation speed-up, parallel fraction of the code and overall efficiency of the parallel implementation are discussed in details.

  13. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

    Directory of Open Access Journals (Sweden)

    Hsoung-Wei Chou

    2015-01-01

    Full Text Available The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

  14. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  15. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  16. Modeling and simulation of high-pressure industrial autoclave polyethylene reactor

    Directory of Open Access Journals (Sweden)

    2008-01-01

    Full Text Available High-pressure technology for polyethylene production has been widely used by industries around the world. A good model for the reactor fluid dynamics is essential to set the operating conditions of an autoclave reactor. The high-pressure autoclave reactor model developed in this work was based on a non-isothermal dynamic model, where PID control equations are used to maintain the operation at the unstable steady state. The kinetic mechanism to describe the polymerization rate and molecular weight averages are presented. The model is capable of computing temperature, concentration gradients and polymer characteristics. The model was validated for an existing industrial reactor and data for production of homopolymer polyethylene and has represented well the behavior of the autoclave reactor used in ethylene homopolymerization.

  17. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  18. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  19. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  20. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  1. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  2. 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

  3. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  4. Dynamic changes of bacterial community in activated sludge with pressurized aeration in a sequencing batch reactor.

    Science.gov (United States)

    Zhang, Yong; Jiang, Wei-Li; Qin, Yang; Wang, Guo-Xiang; Xu, Rui-Xiao; Xie, Biao

    2017-06-01

    This study aimed to investigate the organic removal efficiency and microbial population dynamics in activated sludge with pressurized aeration. The activated sludge was fed with synthetic wastewater composed of simple carbon source to avoid the effect of complex components on microbial communities. The pressurized acclimation process was conducted in a bench-scale sequencing batch reactor (SBR) under 0.3 MPa gage pressure. Another SBR was running in atmospheric environment as a control reactor, with the same operation parameters except for the pressure. Bacterial diversity was investigated by Illumina sequencing technology. The results showed that the total organic carbon removal efficiency of the pressurized reactor was significantly higher, while the mixed liquor suspended solids concentrations were much lower than those of the control reactor. Moderate pressure of 0.3 MPa had little effect on Alpha-diversity of bacterial communities due to the similar running conditions, e.g., feed water, solids retention time (SRT) and the cyclic change of dissolved oxygen (DO) concentrations. Although the relative percentage of the bacterial community changed among samples, there was no major change of predominant bacterial populations between the pressurized group and the control group. Pressurized aeration would have a far-reaching impact on microbial community in activated sludge when treating wastewaters being unfavorable to the dissolution of oxygen.

  5. Sequence of decommissioning of the main equipment in a central type VVER 440 V-230; Secuencia de desmantelamiento de los equipos principales de una central Tipo VVer 440 V-230

    Energy Technology Data Exchange (ETDEWEB)

    Andres, E.; Garcia Ruiz, R.

    2014-10-01

    IBERDROLA Ingenieria y Construccion S.A.U., leader of consortium with Empresarios Agrupados and INDRA, has developed the Basic Engineering for the decommissioning of contaminated systems and building of a VVER 440 V-230 Nuclear Power Plant, establishing the sequence and methodology for the main equipment fragmentation. For that, it has been designed dry and wet cutting zones to be set up in the area where steam generators, main cooling pumps and pressurizer are located; these components will be dismantled previously. (Author)

  6. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  7. Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryson, J.W.; Dickson, T.L.; Malik, S.N.M.; Simonen, F.A.

    1999-08-01

    The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.

  8. The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pin-Chiun [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China); Chou, Hsoung-Wei, E-mail: hwchou@iner.gov.tw [Institute of Nuclear Energy Research, Taoyuan 32546, Taiwan, ROC (China); Ferng, Yuh-Ming [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China)

    2016-02-15

    Highlights: • Probabilistic fracture mechanics method was used to analyze a reactor pressure vessel. • Effects of copper and nickel contents on RPV fracture probability under PTS were investigated and discussed. • Representative PTS transients of Beaver Valley nuclear power plant were utilized. • The range of copper and nickel contents of the RPV materials were suggested. • With different embrittlement levels the dominated PTS category is different. - Abstract: The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

  9. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  10. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    local power distribution within the core. The code's predictions are strongly influenced by the way how coolant mixing is modeled by the analyst by means of 1D-thermal hydraulic codes. Sensitivity evaluations are therefore necessary to identify the most important phenomena and assumptions affecting the numerical predictions of coupled codes. Selected results of these investigations will be presented and discussed. A comparison of the thermal hydraulic data obtained during the tests with the code's predictions will be also given. It can be stated that even though the overall trends of most plant parameters are in a reasonable agreement with the experimental data, these investigations show that multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel. Hence the subsequent Phase 2 of this V1000-CT benchmark is focused on CFD-based simulations for transient conditions typical of a main steam line break transients of VVER-1000 reactors. (orig.)

  11. The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

    Science.gov (United States)

    Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

    2014-10-01

    The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

  12. Structure analysis of a reactor pressure vessel by two- and three-dimensional models. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sacher, H.; Mayr, M.

    1982-03-01

    This paper investigates the reactor pressure vessel of a 1300 MW pressurised water reactor. In order to determine the stresses and deformations of the vessel, two- and three-dimensional finite element models are used which represent the real structure with different degrees of accuracy. The results achieved by these different models are compared for the case of the transient called ''Start up of the nuclear power plant''. 5 refs.

  13. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael Swanson; Daniel Laudal

    2008-03-31

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio

  14. Hydrogen storage using a hot pressure swing reactor

    OpenAIRE

    Preuster, P.; Dürr, S.; Seidel, A.; Müller, K.; Bösmann, A.; Jorschick, Holger; Wasserscheid, Peter

    2017-01-01

    Our contribution demonstrates that hydrogen storage in stationary Liquid Organic Hydrogen Carrier (LOHC) systems becomes much simpler and significantly more efficient if both, the LOHC hydrogenation and the LOHC dehydrogenation reaction are carried out in the same reactor using the same catalyst. The finding that the typical dehydrogenation catalyst for hydrogen release from perhydro dibenzyltoluene (H18-DBT), Pt on alumina, turns into a highly active and very selective dibenzyltoluene hydrog...

  15. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  16. Department of Energy's team's analyses of Soviet designed VVERs

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document provides Appendices A thru K of this report. The topics discussed respectively are: radiation induced embrittlement and annealing of reactor pressure vessel steels; loss of coolant accident blowdown analyses; LOCA blowdown response analyses; non-seismic structural response analyses; seismic analyses; S'' seal integrity; reactor transient analyses; fire protection; aircraft impacts; and boric acid induced corrosion. (FI).

  17. Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg; Pavlova, M., E-mail: pavlova@inrne.bas.bg

    2015-12-15

    Highlights: • We validate operator actions in case of primary to secondary leakage. • We perform four scenarios related to SGTR accident for VVER-1000/V320. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP5/MOD 3.2 computer code is used in performing the analyses. • The analyses confirm the effectiveness of operator actions during PRISE. - Abstract: This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by: • back up filling the ruptured SG; • using letdown system in the affected SG and • by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS). This paper is possible through the participation of leading specialists from KNPP.

  18. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  19. 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.

  20. Physical startup tests for VVER-1200 of Novovoronezh NPP. Advanced technique and some results

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, Dmitry A.; Kraynov, Yury A.; Pinegin, Anatoly A.; Tsyganov, Sergey V. [National Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    2017-09-15

    The intention of the startup physics tests was to confirm design characteristics of the core loading and their compliance with safety analysis preconditions. The program of startup tests for the leading unit is usually composed in such a way that is is possible to study as much neutron-physical characteristics as possible in the safest condition of zero power. State-of-the-art safety analysis is including computer codes that use three dimensional neutron kinetics and thermohydraulics models. For the substantiation of such models, for its validation and verification there is a need in reactor experiments that implementing spatially distributed transients. We based on such statements when composing hot zero power physical startup program for the new VVER-1200 unit of Novovoronezh NPP. Several tests unconventional for VVER were developed for that program. It includes measuring the worth for each of control rod groups and measuring of single rod worth from the inserted groups - test that models rod ejection event in some sense.

  1. An economic optimization of pressurized light water reactor cores

    Science.gov (United States)

    Pfeifer, Holger

    Two reactor cores (1000 MWe and 600 MWe) are optimized with respect to power cost. The power cost is minimized while retaining the thermal-hydraulic margins of the reference core. Constant thermal-hydraulic margins result in similar accident thermal-hydraulic transient behavior of the cores developed during the optimization study. The cost components impacted by the optimization are once-through fuel cycle, capital, and administrative/manpower costs. The variables in the optimization are pin diameter, moderator to fuel (H/U) ratio, core length, and the number of fuel pins in the core. A sequential quadratic programming approach is employed to solve the nonlinear optimization problem with constraints. The fuel cycle costs are evaluated by the use of the linear reactivity model, and capital costs are adjusted by suitable modifications to the nuclear energy cost database reference costs. The results of the analysis shows that for fixed assembly parameters (i.e., pin diameter, H/U ratio, and core length), the optimum core is one that operates at the thermal-hydraulic limits. Cores optimized with unconstrained assembly characteristics contain a larger number of smaller pins at a higher H/U ratio. This follows the trend in current reactor designs. While the lifetime power cost savings for the optimized core are less than 4 million dollars (versus a present day total cost of 6.9 billion dollars), the optimization analysis shows that higher thermal-hydraulic margins can be attained with minimum power cost increases. With increased emphasis on reactor safety, significantly higher safety margins may therefore be achieved without a significant power cost increase. The optimized configurations were found to be relatively insensitive to fuel cycle cost component variations.

  2. 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.

  3. Franco-German nuclear cooperation: from the `common product` to the first European pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vignon, D. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-la-Defense (France)

    1999-01-01

    It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the `common product`, this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales.

  4. 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

  5. A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Sunil Sunny, C. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India); Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-05-01

    Utilization of Mixed Uranium-Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse 'AVVER-1000LEUandMOXAssemblyComputationalBenchmark' and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group 'JEFF31GX' cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k{sub {infinity}}). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium-Gadolinium (UGD) pin, fission rate distributions in UGD, UO{sub 2} and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.

  6. Pressure suppression containment system for boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  7. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  8. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  9. Optimization of Pressurized Oxy-Combustion with Flameless Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malavasi, Massimo [Itea SpA, Gallarate, VA (Italy); Landegger, Gregory [ThermoEnergy Corporation, Worcester, MA (United States)

    2014-06-30

    Pressurized OxyECombustion is one of the most promising technologies for utility-scale power generation plants. Benefits include the ability to burn low rank coal and capture CO2. By increasing the flue gas pressure during this process, greater efficiencies are derived from increased quantity and quality of thermal energy recovery. UPA with modeling support from MIT and testing and data verification by Georgia Tech’s Research Center designed and built a 100 kW system capable of demonstrating pressurized oxyEcombustion using a flameless combustor. Wyoming PRB coal was run at 15 and 32 bar. Additional tests were not completed but sampled data demonstrated the viability of the technology over a broader range of operating pressures, Modeling results illustrated a flat efficiency curve over 20 bar, with optimum efficiency achieved at 29 bar. This resulted in a 33% (HHV) efficiency, a 5 points increase in efficiency versus atmospheric oxy-combustion, and a competitive cost of electricity plus greater CO2 avoidance costs then prior study’s presented. UPA’s operation of the bench-scale system provided evidence that key performance targets were achieved: flue gas sampled at the combustor outlet had non-detectable residual fly ashes, and low levels of SO3 and heavy-metal. These results correspond to prior pressurized oxy-combustion testing completed by IteaEEnel.

  10. Pyrolysis and Combustion of Pulverized Wheat Straw in a Pressurized Entrained Flow Reactor

    DEFF Research Database (Denmark)

    Fjellerup, Jan Søren; Gjernes, Erik; Hansen, Lars Kresten

    1996-01-01

    at relevant conditions. The pressurized entrained now reactor designed at Rise is introduced. Pyrolysis and combustion at 10 and 20 bar pressure have been studied using pulverized wheat straw. Samples of partly reacted particles are collected, and the conversion is calculated using the ash tracer technique....... The pyrolysis experiments show a yield larger than the yield from the proximate analysis. The pyrolysis is completed in about Is, and the yield is the same for 10 and 20 bar pressure. The combustion experiments show a high reactivity with oxygen, and the effects of pressure on combustion are discussed using...... the 10 and 20 bar experiments....

  11. The probabilistic 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 [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Lu, Feng; Wang, Rongshan; Yu, Weiwei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Wang, Donghui [State Nuclear Power Plant Service Company, 200237 Shanghai (China); Zhang, Guodong; Xue, Fei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China)

    2015-12-01

    Highlights: • The methodology and the case study of the FAVOR software were shown. • The over-conservative parameters in the DFM were shown. • The differences between the PFM and the DFM were discussed. • The limits in the current FAVOR were studied. - Abstract: The pressurized thermal shock (PTS) event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel (RPV) during the long time operation (LTO). In the USA, the “screening criteria” for maximum allowable embrittlement of RPV material, which forms part of the USA regulations, is based on the probabilistic fracture mechanics (PFM). The FAVOR software developed by Oak Ridge National Laboratory (ORNL) is used to establish the regulation. As the technical basis of FAVOR is not the most widely-used and codified methodologies, such as the ASME and RCC-M codes, in most countries (with exception of the USA), proving RPV integrity under the PTS load is still based on the deterministic fracture mechanics (DFM). As the maximum nil-ductility-transition temperature (RT{sub NDT}) of the beltline material for the 54 French RPVs after 40 years operation is higher than the critical values in the IAEA-TECDOC-1627 and European NEA/CSNI/R(99)3 reports (while still obviously lower than the “screening criteria” of the USA), it may conclude that the RPV will not be able to run in the LTO based on the DFM. In the FAVOR, the newest developments of fracture mechanics are applied, such as the warm pre-stress (WPS) effect, more accurate estimation of the flaw information and less conservation of the toughness (such as the three-parameter Weibull distribution of the fracture toughness). In this paper, the FAVOR software is first applied to show both the methodology and the results of the PFM, and then the limits in the current FAVOR software (Version 6.1, which represents the baseline for re-assessing the regulation of 10 CFR 50.61), lack of the impact of the constraint effect

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

    Science.gov (United States)

    Hill, Paul R.

    1994-01-01

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

  13. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  14. Heat transport in cold-wall single-wafer low pressure chemical-vapor-deposition reactors

    NARCIS (Netherlands)

    Hasper, A.; Schmitz, J.E.J.; Holleman, J.; Verweij, J.F.

    1992-01-01

    A model is formulated to understand and predict wafer temperatures in a tungsten low pressure chemical‐vapor‐deposition (LPCVD) single‐wafer cold‐wall reactor equipped with hot plate heating. The temperature control is usually carried out on the hot plate temperature. Large differences can occur

  15. Advanced Computational Thermal Studies and their Assessment for Supercritical-Pressure Reactors (SCRs)

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; J. Y. Yoo; J. S. Lee; S. T. Ro; E. Lurien; S. O. Park; R. H. Pletcher; B. L. Smith; P. Vukoslavcevic; J. M. Wallace

    2009-04-01

    The goal of this laboratory / university collaboration of coupled computational and experimental studies is the improvement of predictive methods for supercritical-pressure reactors. The general objective is to develop supporting knowledge needed of advanced computational techniques for the technology development of the concepts and their safety systems.

  16. A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy.

    Science.gov (United States)

    Beach, N J; Knapp, S M M; Landis, C R

    2015-10-01

    The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from -90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor to be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.

  17. Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

    Science.gov (United States)

    Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

    2015-02-01

    Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. 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 and...

  19. Experimental investigation of non-equilibrium thermal-hydraulic processes in a new passive VVER core reflooding system

    Energy Technology Data Exchange (ETDEWEB)

    Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Alexandr A Tzyganok [State Scientific Center of Russian Federation - Institute for Physics and Power Engineering by A.I. Leypunsky, IPPE, 1, Bondarenko sq., Obninsk, 249030 (Russian Federation)

    2005-07-01

    Full text of publication follows: In the systems of passive reactor cooling and core reflooding from above under accident conditions, the cold water flows out into countercurrent steam flow; that is, the direct contact between steam and liquid occurs simultaneously over all cross-sections of pipes/closed spaces. The process is complicated by additional effect of steam condensation and droplet flow towards the steam flow. The processes of heat and mass transfer proceed simultaneously at variable liquid level. These factors give rise to a nonsteady, non-equilibrium two-phase flow. The operating conditions of such a non-steady process, the main of which being the time of closed water volume dump (or the flow coefficient), can be obtained only experimentally. The present work is devoted to the experimental investigations of the interaction of saturated steam with cold water at its flowing out from a closed space with a variable level and some characteristics of dynamic two-phase layer at the steam/liquid interface. With reference to the system of passive heat removal from VVER core, the processes of interaction of saturated steam, steam-water mixture, and air with cold water at its flowing out from a vertical plugged pipe with an internal diameter of 50 and 100 mm have been studied at a pressure of 0.5 MPa. It has been stated experimentally that the dump rate of subcooled water from a plugged pipe into steam is nearly an order of magnitude less than that into non-condensable gas media. The semi-empirical correlation describing the processes of water outflow from plugged pipes into steam is of the form of: W-bar = C{sub 0} {radical}(gd), where C{sub 0} is the dimensionless constant, d is the pipe internal diameter, m, g = 9.81 m/s{sup 2}. (authors)

  20. Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2016-12-15

    Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-15

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

  2. Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

    Energy Technology Data Exchange (ETDEWEB)

    Zaidabadinejad, Majid; Ansarifar, Gholam Reza [Isfahan Univ. (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-11-15

    In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

  3. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

  4. Fatigue crack propagation in steels for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Klesnil, M.; Lukas, P.; Kunz, L. (Ceskoslovenska Akademie Ved, Brno. Ustav Fyzikalni Metalurgie); Troshchenko, V.T.; Pokrovskij, V.V.; Yasnij, P.V.; Skorenko, Y.S. (AN Ukrainskoj SSR, Kiev. Inst. Problem Prochnosti)

    1983-01-01

    Fatigue crack propagation data were measured on 15Kh2NMFA steel of Czechoslovak and Soviet makes. The results obtained by two laboratories were compared with other available data regarding materials for pressure vessels of nuclear power plants. Crack propagation curves were measured at temperatures -60, 20 and 350 degC and the corresponding parameters of crack growth equation were found. Threshold values of stress intensity factor amplitude, Ksub(apz), and the influence of stress ratio R in the range of small crack rates were determined experimentally. Fractography revealed either transgranular or mixed transgranular and interaranular fracture modes depending on stress intensity amplitude Ksub(a) and the environment.

  5. Conditioning of waste from the dismantling of reactor pressure vessel components and of core components

    Energy Technology Data Exchange (ETDEWEB)

    Cleve, R.; Oldiges, O.; Splittler, U.

    2001-07-01

    When power reactors are shut down, the treatment of the core components and of the reactor pressure vessel components constitutes a particular task. Due to its special radiological characteristics, this waste must basically be handled using remote-control equipment with regard to the treatment. For its conditioning in a manner appropriate for the final storage, the waste must be dried and packed into suitable packing drums. The solution implemented by GNS and DSD in order to treat this waste from the Wuergassen nuclear power plant is described in the present contribution. (orig.)

  6. Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  7. Design strategies for optically-accessible, high-temperature, high-pressure reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  8. KARATE - a code for VVER-440 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.

    1994-12-31

    A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

  9. Stress analysis of the reactor pressure vessel of the high performance light water reactors (HPLWR); Festigkeitsanalyse fuer den Reaktordruckbehaelter des High Performance Light Water Reactor (HPLWR)

    Energy Technology Data Exchange (ETDEWEB)

    Guelton, E.; Fischer, K.

    2006-12-15

    The High Performance Light Water Reactor (HPLWR) is one of the concepts of the Generation IV program. The main difference compared to current Light Water Reactors (LWR) results from the supercritical steam condition of the coolant. Due to the supercritical pressure of 25 MPa, water, used as moderator and coolant, flows as a single phase through the core. The temperatures at the outlet are above 500 C. These conditions have a major impact on the design of the Reactor Pressure Vessel (RPV). For the modelling a RPV concept is proposed, which resembles the design of current LWR and allows the use of approved materials on one side and also meets the additional demands on the other side. A first dimensioning of the RPV wall thicknesses and the geometrical proportions has been performed using the german KTA-guidelines. To verify these results, a stress analysis using the finite element method has been performed with the program ANSYS. The combined mechanical and thermal calculations provide the primary, secondary and peak stresses which are evaluated using the KTA-guidelines design loading (Level 0) and service loading level A for the different components. The results confirm the wall thicknesses estimated by Fischer et al. (2006), but there are peak stresses in the vicinity of the inlet and outlet flanges, which are very close to the allowed design limit. For larger diameters of the RPV those regions will become critical and the stresses might exceed the design limits. Design optimizations for those regions are proposed and evaluated. A readjusted geometry of the inlet flange reduces those stresses by 65%. (orig.)

  10. Support device in a pressure vessel, particularly a reactor pressure vessel, to support against horizontal forces. Abstuetzeinrichtung an einem Druckbehaelter, insbesondere einem Reaktordruckbehaelter, gegen Horizontalkraefte

    Energy Technology Data Exchange (ETDEWEB)

    Dorner, H.; Harand, E.; Scholz, M.; Scheler, R.

    1986-04-17

    In a support device on a reactor pressure vessel of a PWR, a guide leading downwards is provided in the vertical pellet cartridge on the floor side of the reactor pressure vessel, which is located in a fixed support pipe. The support pipe is supported on a concreted baseplate in the horizontal direction, and can be screwed to this. The support pipe has a plate-shaped support foot to support it on the baseplate. The reactor pressure vessel, normally resting on its main coolant duct in the plane normal to the axis, can be effectively supported against horizontal forces by this additional support, without preventing thermal movement in the longitudinal axis.

  11. 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.

  12. Combining COMSOL modeling with acoustic pressure maps to design sono-reactors.

    Science.gov (United States)

    Wei, Zongsu; Weavers, Linda K

    2016-07-01

    Scaled-up and economically viable sonochemical systems are critical for increased use of ultrasound in environmental and chemical processing applications. In this study, computational simulations and acoustic pressure maps were used to design a larger-scale sono-reactor containing a multi-stepped ultrasonic horn. Simulations in COMSOL Multiphysics showed ultrasonic waves emitted from the horn neck and tip, generating multiple regions of high acoustic pressure. The volume of these regions surrounding the horn neck were larger compared with those below the horn tip. The simulated acoustic field was verified by acoustic pressure contour maps generated from hydrophone measurements in a plexiglass box filled with water. These acoustic pressure contour maps revealed an asymmetric and discrete distribution of acoustic pressure due to acoustic cavitation, wave interaction, and water movement by ultrasonic irradiation. The acoustic pressure contour maps were consistent with simulation results in terms of the effective scale of cavitation zones (∼ 10 cm and COMSOL simulations. In this study, verification of simulation results with experiments demonstrates that coupling of COMSOL simulations with hydrophone measurements is a simple, effective and reliable scientific method to evaluate reactor designs of ultrasonic systems. Copyright © 2016 Elsevier B.V. All rights reserved.

  13. Cutting and conditioning of the reactor pressure vessel in the NPP Wuergassen; Zerlegung und Konditionierung des Reaktordruckgefaesses im Kernkraftwerk Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Kraps, Uwe [AREVA NP GmbH, Erlangen (Germany); Duwe, Peter [E.ON Kernkraft GmbH, Bewerungen (Germany)

    2011-07-01

    NPP Wuergassen was shutdown in 1995 after 23 years of operation. Since 1997 the nuclear power plant is being dismantled. The cutting of the reactor pressure vessel internals was performed between 2003 and 2008. After decontamination the cylindrical parts of the reactor pressure vessel were dissected, the process was finalized in 2010. AREVA has now a 30 years-experience concerning repair, replacement and dismantling of reactor components. In the contribution the authors describe the process planning, manufacture and testing of appropriate remote handled tools, decontamination, dissection of the pressure vessel (320 t), conditioning, packaging and transport of the radioactive waste including radiation protection monitoring.

  14. 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.

  15. Nuclear power station with a water-cooled reactor pressure vessel. Kernkraftwerk mit einem wassergekuehlten Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, R.; Brunner, G.; Jost, N.

    1987-10-29

    Nuclear radiation produces radiolysis gases, which are undesirable for corrosion and oxyhydrogen gas reasons. To limit the proportion of this radiolysis gas, the invention provides that catalytic surfaces should be introduced into the primary circuit, to produce recombination of hydrogen and oxygen. These surfaces can be accommodated in the upper part of the reactor pressure vessel. The live steam screen can also have a catalytic surface.

  16. Implementation of 3D-Imaging technique for visual testing in a nuclear reactor pressure vessel

    OpenAIRE

    Tanco, André

    2014-01-01

    This master thesis has been performed by request of Dekra Industrial AB. Dekra Industrial AB is a Swedish subsidiary company of the German company Dekra and works for example with safety inspections within the nuclear power industry. The inspections performed by the company are often non-destructive testing (NDT) such as visual inspections of nuclear reactor pressure vessels. The inspection methods used today are considered to be further developed and there is a strong demand of improving the...

  17. The role of water chemistry for environmentally assisted cracking in low-alloy reactor pressure vessel and piping steels under boiling reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2005-07-01

    The environmentally assisted initiation and propagation of cracks in structural materials is one of the most important degradation and ageing mechanisms in light water reactors (LWR) and may seriously affect plant availability and economics. In the first part of this paper a short general introduction on environmentally assisted cracking (EAC) and its significance for LWR is given. Then the important role of water chemistry control in reducing the EAC risk in LWR is illustrated by current research results about the effect of chloride transients and hydrogen water chemistry on the EAC crack growth behaviour of low-alloy reactor pressure vessel and piping steels under boiling water reactor conditions. (author)

  18. 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).

  19. 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

  20. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  1. Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R.; Hiser, A.L. (Materials Engineering Associates, Inc., Lanham, MD (USA))

    1990-03-01

    This report describes a set of experiments undertaken using a 2 MW test reactor, the UBR, to qualify the significance of fluence rate to the extent of embrittlement produced in reactor pressure vessel steels at their service temperature. The test materials included two reference plates (A 302-B, A 533-B steel) and two submerged arc weld deposits (Linde 80, Linde 0091 welding fluxes). Charpy-V (C{sub v}), tension and 0.5T-CT compact specimens were employed for notch ductility, strength and fracture toughness (J-R curve) determinations, respectively. Target fluence rates were 8 {times} 10{sup 10}, 6 {times} 10{sup 11} and 9 {times} 10{sup 12} n/cm{sup 2} {minus}s{sup {minus}1}. Specimen fluences ranged from 0.5 to 3.8 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The data describe a fluence-rate effect which may extend to power reactor surveillance as well as test reactor facilities now in use. The dependence of embrittlement sensitivity on fluence rate appears to differ for plate and weld deposit materials. Relatively good agreement in fluence-rate effects definition was observed among the three test methods. 52 figs., 4 tabs.

  2. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  3. 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

  4. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  5. 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.

  6. On-line Analysis of Catalytic Reaction Products Using a High-Pressure Tandem Micro-reactor GC/MS.

    Science.gov (United States)

    Watanabe, Atsushi; Kim, Young-Min; Hosaka, Akihiko; Watanabe, Chuichi; Teramae, Norio; Ohtani, Hajime; Kim, Seungdo; Park, Young-Kwon; Wang, Kaige; Freeman, Robert R

    2017-01-01

    When a GC/MS system is coupled with a pressurized reactor, the separation efficiency and the retention time are directly affected by the reactor pressure. To keep the GC column flow rate constant irrespective of the reaction pressure, a restrictor capillary tube and an open split interface are attached between the GC injection port and the head of a GC separation column. The capability of the attached modules is demonstrated for the on-line GC/MS analysis of catalytic reaction products of a bio-oil model sample (guaiacol), produced under a pressure of 1 to 3 MPa.

  7. CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

    Directory of Open Access Journals (Sweden)

    I. Spasov

    2010-01-01

    Full Text Available The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT and include asymmetric vessel flow transients and main steam line break (MSLB transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

  8. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  9. 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.

  10. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  11. 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.

  12. Structure analysis of a reactor pressure vessel by two and three-dimensional models

    Energy Technology Data Exchange (ETDEWEB)

    Sacher, H.; Mayr, M. (Technischer Ueberwachungs-Verein Bayern e.V., Muenchen (Germany, F.R.))

    1982-03-01

    This paper investigates the reactor pressure vessel of a 1300 MW pressurised water reactor. In order to determine the stresses and deformations of the vessel, two- and three-dimensional finite element models are used which represent the real structure with different degrees of accuracy. The results achieved by these different models are compared for the case of the transient called 'Start up of the nuclear power plant'. It was found that axisymmetric models, which consider non-axisymmetric components by correction factors, together with special attention to holes and other stress concentrations, allow a sufficient computation of stresses and deformations in the vessel, with the exception of the coolant nozzle region. In this latter case a fully three-dimensional analysis may be necessary.

  13. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto, E-mail: ioliveira@con.ufrj.b, E-mail: schirru@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  14. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  15. Current understanding on the neutron irradiation embrittlement of BWR reactor pressure vessel steels in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Asano, K.; Nishiyama, T. [TEPCO (Japan); Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A. [CRIEPI (Japan); Ohta, T. [Japan Atomic Power Co. (Japan); Ishimaru, Y. [Chugoku EPCO (Japan); Yoneda, H. [Hokuriku EPCO (Japan); Lida, J. [Tohoku EPCO (Japan); Yuya, H. [Chubu EPCO (Japan)

    2011-07-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels has been of concern primarily for the pressurized water reactors (PWRs). After long operation experiences, we are now becoming aware of the situation that the neutron irradiation embrittlement is also of concern for some of the boiling water reactors (BWRs) particularly with Cu-containing RPV steels. The surveillance data of Cu-containing BWR RPV steels show relatively larger shift in ductile-to-brittle transition temperature of fracture toughness than predicted by the embrittlement correlation method developed in late eighties and early nineties. Accurate evaluation of the amount of embrittlement is now very important for long-term operation of BWRs. In this paper, we will describe the neutron irradiation embrittlement of BWR RPVs in Japan. Some of the materials that show relatively large transition temperature shifts are investigated to understand the causes of embrittlement using state-of-the-art microstructural characterization techniques. Furthermore, some archive materials of such RPVs are irradiated in a material testing reactor with high neutron flux to understand the effect of flux on transition temperature shifts and corresponding microstructural changes. Microstructural evolution under irradiation, solute clustering in particular could explain the differences in transition temperature shift of the analyzed specimens. Larger BWR RPVs, which have larger water gaps, receive less neutron irradiation and harmful impurities in steels such as copper are well controlled since 1980 so irradiation embrittlement in BWR vessels can now be considered a concern only in old and small plants. All the new information obtained through these activities was considered in the development of new embrittlement correlation that is now adopted in JEAC 4201- 2007 of Japan Electric Association

  16. 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)

  17. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  18. Study on Pressure drop for Ion Exchanger in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ki-jung; Choi, Jungwoon; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The Jordan Research and Training Reactor (JRTR) is currently being constructed and commissioned in the JUST (Jordan University of Science and Technology) site. The main fluid systems relevant to the JRTR have been proceeding at the Korea Atomic Energy Research Institute. In order to achieve the purpose of the pool water purification, two filters and two ion exchangers which can be to remove suspended solids and ionic impurities in the in-taken pool water have been designed. For the reliable design of this system pump, it is important to predict the pressure drop of the system equipment including the ion exchanger. In this study, the pressure drop in the ion exchanger of PWMS is predicted by using the well-known model and the results provided from manufacturing company. And, the calculated results are compared to the actual data which is measured from the ion exchanger during the PWMS commissioning. The predicted pressure drop is dominated by the resin bed as a portion of about 85% for total pressure drop. The predicted pressure drop is compared to the measured pressure drop of the ion exchanger which is installed in the JRTR, the data above 5 kg/s agree within 5% in the entire range.

  19. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    Energy Technology Data Exchange (ETDEWEB)

    Le Pape, Yann [ORNL; Huang, Hai [Idaho National Laboratory (INL)

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

  20. 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.

  1. Magnetic non-destructive evaluation of hardening of cold rolled reactor pressure vessel steel

    Science.gov (United States)

    Wang, Xuejiao; Qiang, Wenjiang; Shu, Guogang

    2017-08-01

    Non-destructive test (NDT) of reactor pressure vessel (RPV) steel is urgently required due to the life extension program of nuclear power plant. Here magnetic NDT of cold rolled RPV steel is studied. The strength, hardness and coercivity increase with the increasing deformation, and a good linear correlation between the increment of coercivity, hardness and yield strength is found, which may be helpful to develop magnetic NDT of degradation of RPV steel. It is also found that besides dislocation density, the distribution of dislocations may affect coercivity as well.

  2. Welding residual stress distributions for dissimilar metal nozzle butt welds in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Soo; Kim, Ju Hee; Bae, Hong Yeol; OH, Chang Young; Kim, Yun Jae [Korea Univ., Seoul (Korea, Republic of); Lee, Kyungsoo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Song, Tae Kwang [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-02-15

    In pressurized water nuclear reactors, dissimilar metal welds are susceptible to primary water stress corrosion cracking. To access this problem, accurate estimation of welding residual stresses is important. This paper provides general welding residual stress profiles in dissimilar metal nozzle butt welds using finite element analysis. By introducing a simplified shape for dissimilar metal nozzle butt welds, changes in the welding residual stress distribution can be seen using a geometry variable. Based on the results, a welding residual stress profile for dissimilar metal nozzle butt welds is proposed that modifies the existing welding residual stress profile for austenitic pipe butt welds.

  3. 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)

  4. Deformation Characteristics and Sealing Performance of Metallic O-rings for a Reactor Pressure Vessel

    OpenAIRE

    Shen, Mingxue; PENG, Xudong; Xie, Linjun; Meng, Xiangkai; Li, Xinggen

    2016-01-01

    This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the ...

  5. Remote through-wall sampling of the Trawsfynydd reactor pressure vessel: an overview

    Energy Technology Data Exchange (ETDEWEB)

    Curry, A.; Clayton, R. [Magnox Electric, Berkeley (United Kingdom)

    1997-02-01

    This paper summarizes the application of robotic equipment for gaining access to, and removing through-wall samples, from, welds of the reactor pressure vessel at Trawsfyndd power station. The environment, which presents hazards due to ionising radiation, radioactive contamination and asbestos-bearing materials is described. The means of access, by use of remote vehicles with robotic manipulators supported by additional vehicles, it reviewed. The use of abrasive water jet cutting for sample removal is introduced. The relative advantages and disadvantages of this technique are discussed. (Author).

  6. Remote through-wall sampling of the Trawsfynydd reactor pressure vessel: an overview

    Energy Technology Data Exchange (ETDEWEB)

    Curry, A.; Clayton, R. [Magnox Electric, Dartford (United Kingdom). Remote Operations

    1996-12-31

    This paper summarises the application of robotic equipment for gaining access to and removing through-wall samples from welds of the reactor pressure vessel at Trawsfynydd power station. The environment, which presents hazards due to ionising radiation, radioactive contamination and asbestos bearing materials is described. The means of access, by use of remote vehicles complete with robotic manipulators supported by additional vehicles, is reviewed. The use of Abrasive Water Jet Cutting for sample removal is introduced. The relative advantages and disadvantages of this technique are discussed. (UK).

  7. Hydrodynamic analysis of a three-fluidized bed reactor cold flow model for chemical looping hydrogen generation. Pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Xue, Zhipeng; Xiang, Wenguo; Chen, Shiyi; Wang, Dong [Southeast Univ., Nanjing (China). School of Energy and Environment

    2013-07-01

    Chemical looping hydrogen generation (CLHG) can produce pure hydrogen with inherent separation of CO{sub 2} from fossils fuel. The process involves a metal oxide, as an oxygen carrier, such as iron oxide. The CLHG system consists of three reactors: a fuel reactor (FR), a steam reactor (SR) and an air reactor (AR). In the FR, the fuel gases react with iron oxides (hematite Fe{sub 2}O{sub 3}, magnetite Fe{sub 3}O{sub 4}, wuestite FeO), generating reduced iron oxides (FeO or even Fe), and with full conversion of gaseous fuels, pure CO{sub 2} can be obtained after cooling the flue gas from the fuel reactor; in the SR, FeO and Fe reacts with steam to generate magnetite (Fe{sub 3}O{sub 4}) and H{sub 2}, the latter representing the final target product of the process; in the AR, the magnetite is oxidized back to hematite which is used in another cycle. A cold flow model of three-fluidized bed for CLHG corresponding to 50 KW hot units has been built. A major novelty of this facility is the compact fuel reactor, which integrates a bubble and a fast fluidized bed to avoid the incomplete conversion of the fuel gas caused by the thermodynamics equilibrium. In order to study the pressure characteristics and the solids concentration of the system, especially in the fuel reactor, the gas velocity of three reactors, gas flow of L-type value, total solids inventory (TSI) and the secondary air of fuel reactor were varied. Results show that the pressure and the solids concentration are strongly influenced by the fluidizing-gas velocity of three reactors. Moreover, the entrainment of the upper part of fuel reactor increases as the total solids inventory increases, and the operating range of the FR can be changed by introducing secondary air or increasing the total solids inventory.

  8. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  9. Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys

    Energy Technology Data Exchange (ETDEWEB)

    King, J. L. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Jo, H. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Tirawat, R. [National Renewable Energy Laboratory, Concentrating Solar Power Group, Golden, Colorado; Blomstrand, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin; Sridharan, K. [University of Wisconsin–Madison, Department of Engineering Physics, Madison, Wisconsin

    2017-08-31

    Thermal radiation will be an important mode of heat transfer in future high-temperature reactors and in off-normal high-temperature scenarios in present reactors. In this work, spectral directional emissivities of two reactor pressure vessel (RPV) candidate materials were measured at room temperature after exposure to high-temperature air. In the case of SA508 steel, significant increases in emissivity were observed due to oxidation. In the case of Grade 91 steel, only very small increases were observed under the tested conditions. Effects of roughness were also investigated. To study the effects of roughening, unexposed samples of SA508 and Grade 91 steel were roughened via one of either grinding or shot-peening before being measured. Significant increases were observed only in samples having roughness exceeding the roughness expected of RPV surfaces. While the emissivity increases for SA508 from oxidation were indeed significant, the measured emissivity coefficients were below that of values commonly used in heat transfer models. Based on the observed experimental data, recommendations for emissivity inputs for heat transfer simulations are provided.

  10. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  11. Power Maneuvering of Pressurized Water Reactors with Axially Variable Strength Control Rods

    Science.gov (United States)

    Kim, Ung-Soo; Seong, Poong-Hyun

    2004-02-01

    In this research, axially variable strength control rods (AVSCRs) are developed to solve the problems related to the axial power distribution of a reactor during power maneuvering of pressurized water reactors (PWRs). The control rods are classified into two types: multipurpose control rods and regulating control rods. Two multipurpose control rod banks (AVSCR1, AVSCR2) are newly developed; conventional axially uniform strength control rods are adopted as regulating control rod banks to minimize the design change. The newly developed AVSCRs are axially three-sectioned and their worth shapes are optimized to obtain appropriate moving characteristics related to the variation of the axial offset (AO) according to the motion of the AVSCRs. The operation strategy for the power maneuvering is developed in consideration of the moving characteristics of the AVSCRs. This strategy consists of simple logics, and no use of reactivity compensation by boron is considered. Finally, the AVSCRs are applied to the power maneuvering with a typical 100-50-100%, 2-6-2-14 h pattern of daily load-follow for all burn-up state of the core. From the application results, it is shown that the use of AVSCRs makes it possible to regulate AO within the target band during the power maneuvering with only control rods. Consequently, power maneuvering is accomplished without reactivity compensation by a change in boron concentration, and the AVSCRs can cover the entire burn-up states of the reactor core.

  12. 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

  13. Development of Mono-bloc Forging for CAP1400 Reactor Pressure Vessel

    Science.gov (United States)

    Bao-zhong, Wang; Kai-quan, Liu; Ying, Liu; Wen-hui, Zhang; De-li, Zhao

    With the Development of Larger Output and Longer Plant Life for Advanced Pressurized Water Reactor Pressure Vessel (APWRPV), Larger and Mono-bloc Forgings must be Necessary. In the researching and manufacturing of CAP1400RPV forgings, China First Heavy Industries (CFHI) Manufactured Mono-bloc Upper Head with Quick-loc and Mono-bloc Lower Head (Combined Part of Transition Ring and Lower Head),reducing the welding process and the mechanical property deference comparing with the traditional manufacturing, Comparing with AP1000 forgings, not only Obtained better Mechanical Properties, but also Reduced Manufacturing cycle of RPV and Nondestructive in-service Inspection due to Elimination of Weld Seams. CHFI will Development Mono-bloc Nozzle Shell (Combined Part of Vessel Flange Nozzle Shell and Inlet/Outlet Nozzles) for Future APWRPV.

  14. 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.

  15. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  16. Design of a supercritical water-cooled reactor. Pressure vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Kai

    2008-08-15

    The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

  17. Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2014-08-15

    Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

  18. Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

    Directory of Open Access Journals (Sweden)

    Lingyan Zhao

    2014-07-01

    Full Text Available There have been incidents recently where stress corrosion cracking (SCC observed in the dissimilar metal weld (DMW joints connecting the reactor pressure vessel (RPV nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR. Tensile residual stress (RS, in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM, an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

  19. Optical modeling of nickel-base alloys oxidized in pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Clair, A. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France); Foucault, M.; Calonne, O. [Areva ANP, Centre Technique Departement Corrosion-Chimie, 30 Bd de l' industrie, BP 181, 71205 Le Creusot (France); Finot, E., E-mail: Eric.Finot@u-bourgogne.fr [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France)

    2012-10-01

    The knowledge of the aging process involved in the primary water of pressurized water reactor entails investigating a mixed growth mechanism in the corrosion of nickel-base alloys. A mixed growth induces an anionic inner oxide and a cationic diffusion parallel to a dissolution-precipitation process forms the outer zone. The in situ monitoring of the oxidation kinetics requires the modeling of the oxide layer stratification with the full knowledge of the optical constants related to each component. Here, we report the dielectric constants of the alloys 600 and 690 measured by spectroscopic ellipsometry and fitted to a Drude-Lorentz model. A robust optical stratification model was determined using focused ion beam cross-section of thin foils examined by transmission electron microscopy. Dielectric constants of the inner oxide layer depleted in chromium were assimilated to those of the nickel thin film. The optical constants of both the spinels and extern layer were determined. - Highlights: Black-Right-Pointing-Pointer Spectroscopic ellipsometry of Ni-base alloy oxidation in pressurized water reactor Black-Right-Pointing-Pointer Measurements of the dielectric constants of the alloys Black-Right-Pointing-Pointer Optical simulation of the mixed oxidation process using a three stack model Black-Right-Pointing-Pointer Scattered crystallites cationic outer layer; linear Ni-gradient bottom layer Black-Right-Pointing-Pointer Determination of the refractive index of the spinel and the Cr{sub 2}O{sub 3} layers.

  20. Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E., E-mail: thiago.brito86@yahoo.com.br, E-mail: cabol@ufpe.br, E-mail: wagner@unicap.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Universidade Catolica de Pernambuco (UNICAP), Recife, PE (Brazil). Centro de Ciencias e Tecnologia

    2017-11-01

    It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

  1. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    Science.gov (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  2. Characteristics and economy of the European reactor of pressurized water (EPR); Caracteristicas y economia del reactor europeo de agua a presion (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jov@nuclear.inin.mx

    2005-07-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  3. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  4. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  5. Surface Modification of Fine Particle by Plasma Grafting in a Circulating Fluidized Bed Reactor under Reduced Pressure

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sounghee [Woosuk University, Jinchon (Korea, Republic of)

    2015-10-15

    A plasma surface modification of powders has been carried out in a circulating fluidized bed reactor under reduced pressure. Polystyrene (PS) particles treated by plasma are grafted with polyethylene glycol (PEG) on the surface. The virgin, plasma-treated and grafted powders were characterized by DPPH method, FTIR, SEM and contact angle meter. The plasma-treated PS powders have well formed peroxide on the surface, By PEG grafting polymerization, PEG is well grafted and dispersed on the surface of the plasma-treated PS powders. The PEG-g-PS particle was successfully synthesized using the plasma circulating fluidized bed reactor under reduced pressure.

  6. 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.).

  7. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    Energy Technology Data Exchange (ETDEWEB)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l' Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  8. A Neural-Network-Based Nonlinear Adaptive State-Observer for Pressurized Water Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2013-10-01

    Full Text Available Although there have been some severe nuclear accidents such as Three Mile Island (USA, Chernobyl (Ukraine and Fukushima (Japan, nuclear fission energy is still a source of clean energy that can substitute for fossil fuels in a centralized way and in a great amount with commercial availability and economic competitiveness. Since the pressurized water reactor (PWR is the most widely used nuclear fission reactor, its safe, stable and efficient operation is meaningful to the current rebirth of the nuclear fission energy industry. Power-level regulation is an important technique which can deeply affect the operation stability and efficiency of PWRs. Compared with the classical power-level controllers, the advanced power-level regulators could strengthen both the closed-loop stability and control performance by feeding back the internal state-variables. However, not all of the internal state variables of a PWR can be obtained directly by measurements. To implement advanced PWR power-level control law, it is necessary to develop a state-observer to reconstruct the unmeasurable state-variables. Since a PWR is naturally a complex nonlinear system with parameters varying with power-level, fuel burnup, xenon isotope production, control rod worth and etc., it is meaningful to design a nonlinear observer for the PWR with adaptability to system uncertainties. Due to this and the strong learning capability of the multi-layer perceptron (MLP neural network, an MLP-based nonlinear adaptive observer is given for PWRs. Based upon Lyapunov stability theory, it is proved theoretically that this newly-built observer can provide bounded and convergent state-observation. This observer is then applied to the state-observation of a special PWR, i.e., the nuclear heating reactor (NHR, and numerical simulation results not only verify its feasibility but also give the relationship between the observation performance and observer parameters.

  9. Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

    Science.gov (United States)

    Frybort, Jan

    2017-09-01

    Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

  10. Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

    Directory of Open Access Journals (Sweden)

    Frybort Jan

    2017-01-01

    Full Text Available Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

  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. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  13. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant.

  14. Numerical Study of Pollutant Emissions in a Jet Stirred Reactor under Elevated Pressure Lean Premixed Conditions

    Directory of Open Access Journals (Sweden)

    Karim Mazaheri

    2016-01-01

    Full Text Available Numerical study of pollutant emissions (NO and CO in a Jet Stirred Reactor (JSR combustor for methane oxidation under Elevated Pressure Lean Premixed (EPLP conditions is presented. A Detailed Flow-field Simplified Chemistry (DFSC method, a low computational cost method, is employed for predicting NO and CO concentrations. Reynolds Averaged Navier Stokes (RANS equations with species transport equations are solved. Improved-coefficient five-step global mechanisms derived from a new evolutionary-based approach were taken as combustion kinetics. For modeling turbulent flow field, Reynolds Stress Model (RSM, and for turbulence chemistry interactions, finite rate-Eddy dissipation model are employed. Effects of pressure (3, 6.5 bars and inlet temperature (408–573 K over a range of residence time (1.49–3.97 ms are numerically examined. A good agreement between the numerical and experimental distribution of NO and CO was found. The effect of decreasing the operating pressure on NO generation is much more than the effect of increase in the inlet temperature.

  15. High Temperature and Pressure Alkaline Electrochemical Reactor for Conversion of Power to Chemicals

    DEFF Research Database (Denmark)

    Chatzichristodoulou, Christodoulos

    2016-01-01

    Moving away from fossil fuels requires harvesting more and more intermittent renewable energy resources and establishing a sustainable system for the production of chemicals. This brings forward the need for efficient large scale energy storage technologies 1-3 and technologies for the conversion...... of renewable electricity to chemicals. Electrochemical reactors can play a crucial role in this endeavor, since they can efficiently and reversibly transform electricity to high-value chemicals, and thus serve as energy storage and recovery devices for balancing the grid, while offering a means...... for the sustainable production of chemicals 4-6. A novel type of alkaline electrochemical cell that can operate at elevated temperature and pressure has been developed that relies on corrosion resistant high temperature diaphragms, based on mesoporous ceramic membranes where aqueous KOH is immobilized by capillary...

  16. Synthesis of Carbon Nanotubes in Thermal Plasma Reactor at Atmospheric Pressure

    Directory of Open Access Journals (Sweden)

    Lukasz Szymanski

    2017-02-01

    Full Text Available In this paper, a novel approach to the synthesis of the carbon nanotubes (CNTs in reactors operating at atmospheric pressure is presented. Based on the literature and our own research results, the most effective methods of CNT synthesis are investigated. Then, careful selection of reagents for the synthesis process is shown. Thanks to the performed calculations, an optimum composition of gases and the temperature for successful CNT synthesis in the CVD (chemical vapor deposition process can be chosen. The results, having practical significance, may lead to an improvement of nanomaterials synthesis technology. The study can be used to produce CNTs for electrical and electronic equipment (i.e., supercapacitors or cooling radiators. There is also a possibility of using them in medicine for cancer diagnostics and therapy.

  17. 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.

  18. Study of a neutron irradiated reactor pressure vessel steel by X-ray absorption spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)], E-mail: sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2008-11-15

    Reactor pressure vessel (RPV) reference steel samples submitted to neutron irradiations followed by thermal annealing were investigated by X-ray absorption fine structure (XAFS) spectroscopy. Several studies revealed that Cu and Ni impurities can form nanoclusters. In the unirradiated sample and in the only-irradiated sample no significant clustering is detected. In all irradiated and subsequently annealed samples increases of Cu and Ni atom densities are recorded around the absorber. Furthermore, the density of Cu and Ni atoms determined in the first and second shells around the absorber is found to be affected by the irradiation and annealing treatment. The comparison of the XAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the local irradiation damage yields vacancy fractions which were determined from the analysis of XAFS data with a precision of {approx}5%.

  19. Mechanical spectroscopy of reactor-pressure-vessel steel embrittlement: a progress report

    Energy Technology Data Exchange (ETDEWEB)

    Van Ouytsel, K

    1998-08-01

    An enhanced surveillance strategy for testing the fracture toughness of reactor-pressure-vessel steel embrittlement is described. Microstructural investigation in support of damage modelling is an essential element in this enhanced strategy. Temperature-dependent experiments are very sensitive to differences in chemical composition and to effects of neutron irradiation as well as thermal ageing. Amplitude-dependent experiments can be related to tensile test results and correspond to a model for the yield stress. A full range of experiments were carried out on base and weld metal from the Doel-I-II power plants. The results have indicated that internal friction yields information which cannot always be detected by means of standard testing techniques. An inverted torsion pendulum for measuring internal friction has been constructed.

  20. Synthesis of Carbon Nanotubes in Thermal Plasma Reactor at Atmospheric Pressure.

    Science.gov (United States)

    Szymanski, Lukasz; Kolacinski, Zbigniew; Wiak, Slawomir; Raniszewski, Grzegorz; Pietrzak, Lukasz

    2017-02-18

    In this paper, a novel approach to the synthesis of the carbon nanotubes (CNTs) in reactors operating at atmospheric pressure is presented. Based on the literature and our own research results, the most effective methods of CNT synthesis are investigated. Then, careful selection of reagents for the synthesis process is shown. Thanks to the performed calculations, an optimum composition of gases and the temperature for successful CNT synthesis in the CVD (chemical vapor deposition) process can be chosen. The results, having practical significance, may lead to an improvement of nanomaterials synthesis technology. The study can be used to produce CNTs for electrical and electronic equipment (i.e., supercapacitors or cooling radiators). There is also a possibility of using them in medicine for cancer diagnostics and therapy.

  1. A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Simo [Fortum Nuclear Services Ltd., P.O. Box 100, 00048 Fortum (Finland)

    2008-07-01

    Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pin-power reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretized using piecewise linear polynomials. The core is discretized using one node per fuel pin cell. The axial discretization can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modelling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented. HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretizations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. (author)

  2. High Throughput Atomic Layer Deposition Processes: High Pressure Operations, New Reactor Designs, and Novel Metal Processing

    Science.gov (United States)

    Mousa, MoatazBellah Mahmoud

    Atomic Layer Deposition (ALD) is a vapor phase nano-coating process that deposits very uniform and conformal thin film materials with sub-angstrom level thickness control on various substrates. These unique properties made ALD a platform technology for numerous products and applications. However, most of these applications are limited to the lab scale due to the low process throughput relative to the other deposition techniques, which hinders its industrial adoption. In addition to the low throughput, the process development for certain applications usually faces other obstacles, such as: a required new processing mode (e.g., batch vs continuous) or process conditions (e.g., low temperature), absence of an appropriate reactor design for a specific substrate and sometimes the lack of a suitable chemistry. This dissertation studies different aspects of ALD process development for prospect applications in the semiconductor, textiles, and battery industries, as well as novel organic-inorganic hybrid materials. The investigation of a high pressure, low temperature ALD process for metal oxides deposition using multiple process chemistry revealed the vital importance of the gas velocity over the substrate to achieve fast depositions at these challenging processing conditions. Also in this work, two unique high throughput ALD reactor designs are reported. The first is a continuous roll-to-roll ALD reactor for ultra-fast coatings on porous, flexible substrates with very high surface area. While the second reactor is an ALD delivery head that allows for in loco ALD coatings that can be executed under ambient conditions (even outdoors) on large surfaces while still maintaining very high deposition rates. As a proof of concept, part of a parked automobile window was coated using the ALD delivery head. Another process development shown herein is the improvement achieved in the selective synthesis of organic-inorganic materials using an ALD based process called sequential vapor

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

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States)] [and others

    1997-01-01

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

  4. Experimental study of a cocurrent upflow packed bed bubble column reactor: pressure drop, holdup and interfacial area

    NARCIS (Netherlands)

    Molga, E.J.; Westerterp, K.R.

    1997-01-01

    Gas¿liquid interfacial areas have been determined by means of chemically enhanced absorption of CO2 into DEA in a packed bed bubble column reactor with an inner diameter of 156 mm. The influence of the gas velocity and particle diameter on the interfacial areas, pressure drops and liquid holdups has

  5. Transition between trickle flow and pulse flow in a cocurrent gas-liquid trickle-bed reactor at elevated pressures

    NARCIS (Netherlands)

    Wammes, W.J.A.; Mechielsen, S.J.; Westerterp, K.R.

    1992-01-01

    The effect of reactor pressure in the range of 0.2–2.0 MPa on the transition between the trickle-flow and the pulse-flow regime has been investigated for the non-foaming water—nitrogen and aqueous 40% ethyleneglycol—nitrogen systems. Most models and flow charts which are all based on atmospheric

  6. Residence time distribution and hold-up in a cocurrent upflow packed bed reactor at elevated pressure

    NARCIS (Netherlands)

    van Gelder, Klaas B.; Westerterp, K.R.

    1990-01-01

    The residence time distribution in liquid phase was measured in a cocurrent upflow packed bed reactor for the system methanol-hydrogen at low Reynolds numbers and at elevated pressure. The plug flow with axial dispersion model was used to describe mixing in the system. The imperfect pulse method was

  7. Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...

  8. 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...

  9. 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.

  10. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  11. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  12. IVO participation in IAEA benchmark for VVER-type nuclear power plants seismic analysis and testing

    Energy Technology Data Exchange (ETDEWEB)

    Varpasuo, P.

    1997-12-01

    This study is a part of the IAEA coordinated research program `Benchmark study for the Seismic Analysis and Testing of VVER Type NPPs`. The study reports the numerical simulation of the blast test for Paks and Kozloduy nuclear power plants beginning from the recorded free-field response and computing the structural response at various points inside the reactor building. The full-scale blast tests of the Paks and Kozloduy NPPs took place in December 1994 and in July 1996. During the tests the plants operated normally. The instrumentation for the tests consisted of 52 recording channels with 200 Hz sampling rate. Detonating 100 kg charges in 50-meter deep boreholes at 2.5-km distance from the plant carried out the blast tests. The 3D structural models for both reactor buildings were analyzed in the frequency domain. The number of modes extracted in both cases was about 500 and the cut-off frequency was 25 Hz. In the response history run the responses of the selected points were evaluated. The input values for response history run were the three components of the excitation, which were transformed from time domain to the frequency domain with the aid of Fourier transform. The analysis was carried out in frequency domain and responses were transferred back to time domain with inverse Fourier transform. The Paks and Kozloduy blast tests produced a wealth of information on the behavior of the nuclear power plant structures excited by blast type loads containing also the low frequency wave train if albeit with small energy content. The comparison of measured and calculated results gave information about the suitability of the selected analysis approach for the investigated blast type loading. 25 refs.

  13. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  14. 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.

  15. Models for 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. [California Univ., Santa Barbara, CA (United States)

    1995-05-01

    The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  16. Core damage severity evaluation for pressurized water reactors by artificial intelligence methods

    Science.gov (United States)

    Mironidis, Anastasios Pantelis

    1998-12-01

    During the course of nuclear power evolution, accidents have occurred. However, in the western world, none of them had a severe impact on the public because of the design features of nuclear plants. In nuclear reactors, barriers constitute physical obstacles to uncontrolled fission product releases. These barriers are an important factor in safety analysis. During an accident, reactor safety systems become actuated to prevent the barriers from been breached. In addition, operators are required to take specified actions, meticulously depicted in emergency response procedures. In an accident, on-the-spot knowledge regarding the condition of the core is necessary. In order to make the right decisions toward mitigating the accident severity and its consequences, we need to know the status of the core [1, 3]. However, power plant instrumentation that can provide a direct indication of the status of the core during the time when core damage is a potential outcome, does not exist. Moreover, the information from instruments may have large uncertainty of various types. Thus, a very strong potential for misinterpreting incoming information exists. This research endeavor addresses the problem of evaluating the core damage severity of a Pressurized Water Reactor during a transient or an accident. An expert system has been constructed, that incorporates knowledge and reasoning of human experts. The expert system's inference engine receives incoming plant data that originate in the plethora of core-related instruments. Its knowledge base relies on several massive, multivariate fuzzy logic rule-sets, coupled with several artificial neural networks. These mathematical models have encoded information that defines possible core states, based on correlations of parameter values. The inference process classifies the core as intact, or as experiencing clad damage and/or core melting. If the system detects a form of core damage, a quantification procedure will provide a numerical

  17. High pressure water abrasive suspension JET cutting - An innovative cutting technology for the dismantling of reactor core components

    Energy Technology Data Exchange (ETDEWEB)

    Kalwa, H. [VAK, Kahl (Germany); Schwarz, T. [RWE NUKEM Limited, B7 Windscale, Seascale, Cumbria CA20 1PF (United Kingdom)

    2003-07-01

    In the frame of decommissioning of nuclear facilities the dismantling of reactor pressure vessels and their internals represent one special challenge. Due to their high activation, associated high dose rate levels, and to some extent the complex geometry and high material thickness of the components, there are particular demands for dismantling techniques. The task is to safely and economically work in every respect and therefore employ techniques with a wide area of application. As a proven technique, RWE NUKEM offers High Pressure Water Abrasive Suspension Jet Cutting. High pressure Water Abrasive Suspension Jet cutting (WASJ), well established in non-nuclear applications, has now been upgraded to meet the demands of decommissioning in the nuclear industry. The application at the Nuclear Power Plant in Kahl (VAK) was one of the first industrial scale applications. Based on several tests and parametric studies, High Water Abrasive Suspension Jet Cutting was tested against other cutting technologies. Because the overall performance in terms of fast and easy cutting operations, ability for remote handling, production of secondary waste WASJ was chosen at VAK Kahl for the dismantling of the lower core shroud and the reactor pressure vessel itself. The dismantling of the core shroud and the reactor vessel took place in-situ (component in its built-in position) using the WASJ technology. As example of application the core shroud of VAK is given. The total mass of the VAK lower shroud was about 3 tons and the wall thickness varied from 30 to 135 mm. The shroud was cut into segments in its in-vessel position, each segment being 500 x 900 mm and having a mass of about 0.25 tons. Cutting was performed in such a way that the separated pieces could be loaded directly into standard waste containers. All secondary waste (abrasives and dross) was collected in two 200 liter drums and, after drying, the drums were sent directly to waste storage. Reactor Pressure Vessel of VAK

  18. Studying the operation of a VVER steam generator in the condensing mode at different parameters of emergency processes

    Science.gov (United States)

    Morozov, A. V.; Shlepkin, A. S.; Kalyakin, D. S.; Soshkina, A. S.

    2017-05-01

    The article presents the results of the experimental study of heat and mass transfer processes in an NPP steam generator during the operation of passive safety systems of new-generation VVER reactor installations. At the GE2M-PG test rig in the Leypunsky Institute for Physics and Power Engineering, two series of experiments corresponding to different stages of the accident were completed. In these experiments, the performance of VVER steam generator in the condensing mode with and without the removal of gas-vapor mixture from the "cold" header has been studied. As a result of the first series of experiments, it was found that, for any of the parameters of the emergency process, the steam generator's power does not drop below 80% of the original value. Furthermore, we revealed that the composition and physical properties of gases in the investigated concentration range did not notably affect the processes in the steam generator. In the second series of experiments without removal of noncondensable gases, the influence of parameters of the emergency process on the efficiency of heat transfer in the steam generator operating in the condensing mode was investigated. In order to study the heat transfer processes, we studied the change of the temperature difference between the media of the first and second circuits in our experiments. We found that the value of the temperature difference depends on both the mass of noncondensable gases accumulated in the tube bundle and their accumulation rate. The accumulation rate is determined by the power of the steam generator and the concentration of gases entering the steam generator. As a result of the analysis of experimental data, we obtained the analytical dependence reflecting change in the power of the steam generator operating in the emergency condensing mode.

  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. The aging of pressurized water nuclear reactor vessels; Le vieillissement des cuves de reacteurs nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Monin, L. [Autorite de Surete Nucleaire, Dir. des Equipements sous Pression Nucleaires, 75 - Paris (France); Monnot, B. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Bureau d' Analyse des Materiels Mecaniques, 92 - Clamart (France)

    2009-07-15

    In pressurized water nuclear power plants, heat is produced by the fission of the uranium cores, a constituent of the fuel rods placed in the reactor core. This core formed by the set of fuel assemblies is contained In the reactor vessel. As a part of the second confinement barrier of the radioactive elements, the vessel enables the reactor core to be cooled the primary fluid, to be controlled by the control rods, and to be supervised. Its role is of prime importance for the safety of the plant. Its integrity must therefore be guaranteed and must be demonstrated under all operating conditions and for the entire duration of its operation. Contrary to other devices of the primary circuit, like the steam generators, the replacement of a vessel is not an operation considered by EDF. The working life of the facility is consequently linked to the justification of the suitability of the vessel's use. (author)

  1. SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, William M.; Riley, Matthew E.; Spencer, Benjamin W.

    2016-07-01

    In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in

  2. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    Energy Technology Data Exchange (ETDEWEB)

    Jelinek, Tomas

    2001-02-01

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

  3. Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

    2016-04-15

    Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

  4. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  5. 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

  6. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  7. Design of PIλDμ controller for global power control of Pressurized Heavy Water Reactor.

    Science.gov (United States)

    Bongulwar, M R; Patre, B M

    2017-07-01

    In this paper, a robust stabilizing controller design method is presented for global power control of a Pressurized Heavy Water Reactor (PHWR) under step-back condition scheme using a Fractional Order Proportional Integral Derivative (PIλDμ) controller resulting into robust performance. The method is applicable to design a controller for One Non Integer Order Plus Time Delay (NIOPTD-I) plant which satisfies design specifications such as phase margin and gain crossover frequency. Stability boundary locus method is used in (Kp, Ki, Kd) parameter space for NIOPTD-I plants to obtain stability region. The robust performance is obtained by satisfying flat phase condition at gain crossover frequency where phase is almost constant for large span of frequencies. The simulation result of the proposed PIλDμ controller shows active step-back control to the insertion of the rod with no undershoot and with the robust performance, hence safe to the plant for gain variations from 500% lower side to 1000% upper side. The PIλDμ controller with a plant shows that 30% and 50% global power drop from initial 100% is achieved in a reasonable time without undershoot. Copyright © 2017 ISA. Published by Elsevier Ltd. All rights reserved.

  8. Characterization of Nanostructural Features in Irradiated Reactor Pressure Vessel Model Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, B D; Odette, G R; Asoka-Kumar, P; Howell, R H; Sterne, P A

    2001-08-01

    Irradiation embrittlement in nuclear reactor pressure vessel steels results from the formation of a high number density of nanometer-sized copper rich precipitates and sub-nanometer defect-solute clusters. We present results of small angle neutron scattering (SANS) and positron annihilation spectroscopy (PAS) characterization of the nanostructural features formed in binary and ternary Fe-Cu-Mn alloys irradiated at {approx}290 C. These complementary techniques provide insight into the composition and character of both types of nanoscale features. The SANS measurements indicate populations of copper-manganese precipitates and smaller vacancy-copper-manganese clusters. The PAS characterization, including both Doppler broadening and positron lifetime measurements, indicates the presence of essentially defect-free Cu precipitates in the Fe-Cu-Mn alloy and vacancy-copper clusters in the Fe-Cu alloy. Thus the SANS and PAS provide a self-consistent picture of nanostructures composed of copper-rich precipitates and vacancy solute cluster complexes and tend to discount high Fe concentrations in the CRPs.

  9. Development of Abnormal Operating Strategies for Station Blackout in Shutdown Operating Mode in Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hwang, Su-Hyun [FNC Tech. Co., Yongin (Korea, Republic of)

    2016-10-15

    Loss of all AC power is classified as one of multiple failure accident by regulatory guide of Korean accident management program. Therefore we need develop strategies for the abnormal operating procedure both of power operating and shutdown mode. This paper developed abnormal operating guideline for loss of all AC power by analysis of accident scenario in pressurized water reactor. This paper analyzed the loss of ultimate heat sink (LOUHS) in shutdown operating mode and developed the operating strategy of the abnormal procedure. Also we performed the analysis of limiting scenarios that operator actions are not taken in shutdown LOUHS. Therefore, we verified the plant behavior and decided operator action to taken in time in order to protect the fuel of core with safety. From the analysis results of LOUHS, the fuel of core maintained without core uncovery for 73 minutes respectively for opened RCS states after the SBO occurred. Therefore, operator action for the emergency are required to take in 73 minutes for opened RCS state. Strategy is to cooldown by using spent fuel pool cooling system. This method required to change the plant design in some plant. In RCS boundary closed state, first abnormal operating strategy in shutdown LOUHS is first abnormal operating strategy in shutdown LOUHS is to remove the residual heat of core by steam dump flow and auxiliary feedwater of SG.

  10. 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.

  11. Deformation characteristics and sealing performance of metallic-O-ring for a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ming Xue; Peng, Xudong; Xie, Linjun; Meng, Xiang Kai [Engineering Research Center of Process Equipment and Its Remanufacture, Ministry of Education, Zhejiang University of Technology, Hangzhou (China); Li, Xing Gen [Ningbo Tiansheng Sealing Packing Co., Ltd., Ningbo (China)

    2016-04-15

    This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap) have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  12. Deformation Characteristics and Sealing Performance of Metallic O-rings for a Reactor Pressure Vessel

    Directory of Open Access Journals (Sweden)

    Mingxue Shen

    2016-04-01

    Full Text Available This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV. A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  13. 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.

  14. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    Energy Technology Data Exchange (ETDEWEB)

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  15. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  16. 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.

  17. 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.

  18. Phenomenology involved in self-pressurized, natural circulation, low thermo-dynamic quality, nuclear reactors: The thermal-hydraulics of the CAREM-25 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marcel, C.P., E-mail: christian.marcel@cab.cnea.gov.ar [Instituto Balseiro, 8400 S. C de Bariloche (Argentina); Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Furci, H.F.; Delmastro, D.F. [Instituto Balseiro, 8400 S. C de Bariloche (Argentina); Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Masson, V.P. [Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Low quality, natural circulation, and self-pressurized nuclear reactors are modeled analytically. Black-Right-Pointing-Pointer The feedbacks resulting from the interplay of the acting phenomena are analyzed. Black-Right-Pointing-Pointer By decreasing the nuclear power the core inlet enthalpy increases. Black-Right-Pointing-Pointer The mass flow has to be regulated to set it within a certain range in order. - Abstract: The interwoven phenomena involved in a prototypical self-pressurized natural circulation, low thermo-dynamical quality nuclear reactor such as CAREM-25 are analytically presented. These phenomena present many differences with traditional light water nuclear power plants. The dependence between mass flow and core inlet enthalpy on generated power is found. The need of tuning the mass flow rate in accordance to the design value is found to be important in order to keep the thermal margin and the heat transfer coefficients in the steam generators. The influence of condensation in structures or walls in the upper dome on the two-phase boundary is also studied. The dynamic consequences of all these results are therefore discussed. A numerical code is then used to verify the aforementioned findings and to test the validity of the modeling approximations. From the results it is clear that the way the phenomena interact causes the resulting dynamics in CAREM-25 to be substantially different from that existing in reactors such as PWRs, BWRs and also natural circulation BWRs. It is thus clear that the combination of different effects makes CAREM-25 behavior impossible to be extrapolated from existing knowledge and accumulated experience.

  19. An improvement in the use of plutonium in pressurized water reactors; The subassembly mixed-oxide fuel management concept

    Energy Technology Data Exchange (ETDEWEB)

    Bangil, C.; Gambier, G.; Soldevila, M. (Electricite de France DER 1, av. du General de Gaulle 92141 Clamart Cedex (FR))

    1989-09-01

    An unconventional mixed-oxide (MOX) fuel management concept for a pressurized water reactor core is proposed. Subassembly management (SAM) is an in-out reloading scheme applied within MOX assemblies that are reconstructed'' at each end of cycle. The SAM concept needs only one kind of MOX fuel pin instead of the three pins necessary in the standard MOX fuel management concept; it also leads to better use of plutonium.

  20. A research on the Extraction of Metals from Küre Chalcopyrite Concentrate in the Pressure Reactor System

    OpenAIRE

    Hasan Arslanoğlu; Hamdi Soner Altundoğan

    2013-01-01

    In this study, ammonium persulfate leaching of metals from Küre chalcopyrite concentrate was investigated in the pressure reactor by using response surface methodology (RSM). Experiments were planned with all factorial array of central composite design (CCD) as total 86 experiments, so that 10 experiments in centre. The all of investigating parameters having affected on the results were defined as independent variable. The investigating parameters that effect on the metal extraction yield in ...

  1. Dismantling of the reactor pressure vessel internals in the NPP Wuergassen; Zerlegung der RDG- Einbauten im KKW Wuergassen

    Energy Technology Data Exchange (ETDEWEB)

    Bruhn, Jan Hendrik [AREVA NP GmbH (Germany)

    2008-07-01

    The reactor pressure vessel internals of the NPP Wuergassen were dismantled and dissected in the time period 19-03-2007 to 21-09-2007 and packaged for final disposal. The total amount was about 18 tons of steel. The dissected was performed using specific water-quartz sand-cutting tools within a settling tank. The dismantling concept intended to cut shape-optimized pieces with respect to space saving packaging. The project included a specific water treatment system.

  2. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... reactors where neutron radiation has reduced the fracture toughness of the reactor vessel materials, a... operation using appropriate test data. (iii) The methods, including heat source, instrumentation and... monitoring, inspections, and tests proposed to demonstrate that the limitations on temperatures, times and...

  3. Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Seung Hyuk; Ryu, Tae Young; Park, Seung Hyun; Won, Min Gu; Kang, Seok Jun; Kim, Moon Ki; Choi, Jae Boong [Sungkyunkwan University, Suwon (Korea, Republic of); Lee, Kyoung Soo; Lee, Sung Ho [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2015-03-15

    Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

  4. CATALYTIC COMBUSTION OF PROPANE IN A MEMBRANE REACTOR WITH SEPARATE FEED OF REACTANTS .2. OPERATION IN PRESENCE OF TRANS-MEMBRANE PRESSURE-GRADIENTS

    NARCIS (Netherlands)

    SARACCO, G; VELDSINK, JW; VERSTEEG, GF; VANSWAAIJ, WPM

    This is the second communication of a series dealing with an experimental and modelling study on propane catalytic combustion in a membrane reactor with separate feed of reactants. In paper I the behaviour of the reactor in the absence of trans-membrane pressure gradients was presented and

  5. Catalytic combustion of propane in a membrane reactor with separate feed of reactants II. Operation in presence of transmembrane pressure gradients

    NARCIS (Netherlands)

    Saracco, Guido; Veldsink, J.W.; Veldsink, Jan Willem; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1995-01-01

    This is the second communication of a series dealing with an experimental and modelling study on propane catalytic combustion in a membrane reactor with separate feed of reactants. In paper I the behaviour of the reactor in the absence of trans-membrane pressure gradients was presented and

  6. Catalytic combustion of propane in a membrane reactor with separate feed of reactants—II. Operation in presence of trans-membrane pressure gradients

    NARCIS (Netherlands)

    Saracco, Guido; Veldsink, Jan Willem; Versteeg, Geert F.; Swaaij, Wim P.M. van

    1995-01-01

    This is the second communication of a series dealing with an experimental and modelling study on propane catalytic combustion in a membrane reactor with separate feed of reactants. In paper I the behaviour of the reactor in the absence of trans-membrane pressure gradients was presented and

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Li, C.W.; Han, L.Z.; Luo, X.M.; Liu, Q.D.; Gu, J.F., E-mail: gujf@sjtu.edu.cn

    2016-08-15

    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 Fe{sub 3}C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo{sub 2}C 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. - Highlights: • The dependence of the deterioration of impact toughness on tempering temperature has been analysed. • The instrumented Charpy V-notch impact test has been employed to study the fracture mechanism. • The influence of M/A constituents on different fracture mechanisms based on the hinge model has been demonstrated. • A correlation between the mechanical properties and the amount of M/A constituents has been established.

  9. Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

    Science.gov (United States)

    Reid, Terry V.

    2016-01-01

    Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S?A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ?0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

  10. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-12-15

    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  11. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  12. Nuclear reactor neutron shielding

    Science.gov (United States)

    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.

  13. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system.

    Science.gov (United States)

    Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

    2013-01-01

    Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ(13)Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate.

  14. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

    Directory of Open Access Journals (Sweden)

    Yoko eOhtomo

    2013-12-01

    Full Text Available Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in situ pressure (0–100 MPa and temperature (0–70°C conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 mL/min, respectively were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to

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

    Energy Technology Data Exchange (ETDEWEB)

    Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P. [Institute of Physics and Power Engineering State Scientific Center, Obninsk (Russian Federation)

    2001-07-01

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

  16. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  17. Flow Reactor Studies with Nanosecond Pulsed Discharges at Atmospheric Pressure and Higher

    Science.gov (United States)

    2013-10-01

    Reactor Experiments: • Experiments on the oxidation of C1-C7 alkanes with and without plasma assisted reaction performed for T = 300 – 1250 K and P = 1...online) and Gas Chromatography (offline) Inlet Outlet Experimental Facility – cont’d • FTIR (online) connected to exhaust stream of reactor...molecule/pulse Major Species + Minor Species Oxygen + Excited Species C1-C7 Alkanes Plasma Assisted Oxidation Fourth Annual Review Meeting of

  18. Implementation of strength pareto evolutionary algorithm II in the multiobjective burnable poison placement optimization of KWU pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gharari, Rahman [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Poursalehi, Navid; Abbasi, Mohmmadreza; Aghale, Mahdi [Nuclear Engineering Dept, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (K-e-f-f) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

  19. 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.

  20. Implementation of Strength Pareto Evolutionary Algorithm II in the Multiobjective Burnable Poison Placement Optimization of KWU Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Rahman Gharari

    2016-10-01

    Full Text Available In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II, is developed for the burnable poison placement (BPP optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (Keff for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

  1. A research on the Extraction of Metals from Küre Chalcopyrite Concentrate in the Pressure Reactor System

    Directory of Open Access Journals (Sweden)

    Hasan Arslanoğlu

    2013-06-01

    Full Text Available In this study, ammonium persulfate leaching of metals from Küre chalcopyrite concentrate was investigated in the pressure reactor by using response surface methodology (RSM. Experiments were planned with all factorial array of central composite design (CCD as total 86 experiments, so that 10 experiments in centre. The all of investigating parameters having affected on the results were defined as independent variable. The investigating parameters that effect on the metal extraction yield in the chalcopyrite leaching are concentration of ammonium persulfate (APS, leaching temperature, stirring speed and ratio of liquid-solid, reactor occupancy and leaching time. Leaching experiments were performed as batch in the 300 ml pressure reactor. The data obtained from different experimental conditions were optimized that optimizing criteria is maximum copper and minimum iron passing into leach solution, and model equation were formed for both response values. The obtained model equations for copper and iron extraction were determined as second order model. Under the optimum leaching conditions, some model solution points were determined. According to results of performed experiments in these solution points, copper extraction yield is 55% while iron is about 15%.

  2. The application of mechanical and thermal cutting tools for the dismantling of activated internals of the reactor pressure vessels in the Versuchsatomkraftwerk, Kahl and the Gundremmingen Unit A

    Energy Technology Data Exchange (ETDEWEB)

    Eickelpasch, N. [Versuchsatomkraftwerk GmbH, Kahl am Main (Germany); Kalwa, H. [Versuchsatomkraftwerk GmbH, Kahl am Main (Germany); Steiner, H. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, D-89355 Gundremmingen (Germany); Priesmeyer, U. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, D-89355 Gundremmingen (Germany)

    1997-07-01

    The Gundremmingen Unit A plant (KRB A) and the Versuchsatomkraftwerk Kahl (VAK) plant represent the first generation of nuclear reactors in Germany. The 250 MW{sub e} reactor KRB A was the first commercial reactor in Germany and the 16 MW{sub e} reactor VAK was the pilot nuclear power plant, which had to serve mainly scientific purposes. KRB A is under dismantling since 1983, VAK since 1988. Although they are both of the boiling water type, they are rather different to each other, referring to their size and construction. The actual work is the dismantling of high contaminated components inside the reactor buildings and the underwater cutting of activated internals of the reactor pressure vessels. Several cutting techniques have been developed, tested and applied to respective dismantling tasks in the meantime. The experiences made in both projects are not limited to dismantling work only, but also include know-how on effective decontamination and scrap recycling. (orig.)

  3. Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

    Energy Technology Data Exchange (ETDEWEB)

    Hoye, Robert L. Z., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk; MacManus-Driscoll, Judith L., E-mail: rlzh2@cam.ac.uk, E-mail: jld35@cam.ac.uk [Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Muñoz-Rojas, David [LMGP, University Grenoble-Alpes, CNRS, F-3800 Grenoble (France); Nelson, Shelby F. [Kodak Research Laboratories, Eastman Kodak Company, Rochester, New York 14650 (United States); Illiberi, Andrea; Poodt, Paul [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Roozeboom, Fred [Holst Centre/TNO Thin Film Technology, Eindhoven, 5656 AE (Netherlands); Department of Applied Physics, Eindhoven University of Technology, P.O. Box 513, Eindhoven, 5600 MB (Netherlands)

    2015-04-01

    Atmospheric pressure spatial atomic layer deposition (AP-SALD) has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

  4. Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

    Directory of Open Access Journals (Sweden)

    Robert L. Z. Hoye

    2015-04-01

    Full Text Available Atmospheric pressure spatial atomic layer deposition (AP-SALD has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

  5. Pressurized pyrolysis of dried distillers grains with solubles and canola seed press cake in a fixed-bed reactor.

    Science.gov (United States)

    Ateş, Funda; Miskolczi, Norbert; Saricaoğlu, Beyza

    2015-02-01

    Pressurized pyrolysis of biomasses was carried in a fixed bed reactor to obtain gases, bio-oils and chars at elevated temperatures. The products were characterized by GC-MS, FTIR, viscometer, SEM, BET and EDXRFS methods. Experiments were performed at 1, 5 and 10 bar pressure and 400, 500 and 600°C temperatures. The experimental results show that in all the experimental condition the yield of bio-oil from DDGS as higher than that of canola. Yield of non-condensable gases and chars increased, while that of liquid products decreased by pressure. Increasing pressure favoured the formation of low molecular weight gas, such as H2. Maximum surface area of chars was obtained at atmospheric pressure and the surface areas decreased rapidly with increasing pressure. GC/MS results shows that the amount of fatty acids in bio-oils was increased by increasing pressure and bio-oils showed non-Newtonian behavior. Based on EDXRFS results, bio-oils and char contained lots of elements. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL; Terry, Totemeier [Idaho National Laboratory (INL)

    2006-10-01

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

  7. The Low Temperature Oxidation of 2,7-Dimethyloctane in a Pressurized Flow Reactor

    Science.gov (United States)

    Farid, Farinaz

    The complexity of real fuels has fostered the use of simple mixtures of hydrocarbons whose combustion behavior approximates that of real fuels in both experimental and computational studies to develop models of the combustion of the real fuel. These simple mixtures have been called surrogates. Lightly branched paraffins are an important class of constituents in gasoline, diesel and aviation turbine fuels and therefore are primary candidates for use as a component in a surrogate. Unfortunately, fundamental studies on combustion characteristics of high molecular weight mono- and di-methylated iso-paraffins are scarce. Therefore, this study was designed to investigate the low-temperature oxidation of 2,7-dimethyloctane (2,7-DMO) (C10H22), a lightly branched isomer of decane. Replicate 2,7-DMO oxidation experiments were conducted in a pressurized flow reactor (PFR) over the temperature range of 550 -- 850 K, at a pressure of 8 atm and an equivalence ratio of 0.3 in 4.21% oxygen / nitrogen. The reactivity was mapped by continuous monitoring of CO, CO 2, and O2 using a non-dispersive infrared (NDIR) carbon monoxide / carbon dioxide analyzer and an electrochemical oxygen sensor. For examining the underlying reaction chemistry, detailed speciation of samples was performed at selected temperatures using a gas chromatograph with a flame ionization detector coupled to a mass spectrometer. Comparable oxidation experiments for n-decane were carried out to examine the unique effects of branching on fuel reactivity and distribution of major stable intermediates. For both isomers, the onset of negative temperature coefficient (NTC) region was observed near 700 K, with the reactivity decreasing with increasing the temperature. The flow reactor study of n-decane oxidation confirmed that the isomerization reduces the amount of CO produced at peak reactivity. In addition to reaction inhibition, branching affected the distribution of C2-C 4 olefin intermediates. While the oxidation of

  8. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  9. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  10. Synthesis of Fused Pyrimidinone and Quinolone Derivatives in an Automated High-Temperature and High-Pressure Flow Reactor.

    Science.gov (United States)

    Tsoung, Jennifer; Bogdan, Andrew R; Kantor, Stanislaw; Wang, Ying; Charaschanya, Manwika; Djuric, Stevan W

    2017-01-20

    Fused pyrimidinone and quinolone derivatives that are of potential interest to pharmaceutical research were synthesized within minutes in up to 96% yield in an automated Phoenix high-temperature and high-pressure continuous flow reactor. Heterocyclic scaffolds that are either hard to synthesize or require multisteps are readily accessible using a common set of reaction conditions. The use of low-boiling solvents along with the high conversions of these reactions allowed for facile workup and isolation. The methods reported herein are highly amenable for fast and efficient heterocycle synthesis as well as compound scale-ups.

  11. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    CERN Document Server

    Lewis, M E

    2000-01-01

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  12. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  13. Effect of different carbon sources on the biological phosphorus removal by a sequencing batch reactor using pressurized pure oxygen.

    Science.gov (United States)

    Wei, Jie; Imai, Tsuyoshi; Higuchi, Takaya; Arfarita, Novi; Yamamoto, Koichi; Sekine, Masahiko; Kanno, Ariyo

    2014-05-04

    The effect of different carbon source on the efficiency of enhanced biological phosphorus removal (EBPR) from synthetic wastewater with acetate and two ratios of acetate/starch as a carbon source was investigated. Three pressurized pure oxygen sequencing batch reactor (POSBR) experiments were operated. The reactors (POSBR1, POSBR2 and POSBR3) were developed and studied at different carbon source ratios of 100% acetate, 75% acetate plus 25% starch and 50% acetate plus 50% starch, respectively. The results showed that POSBR1 had a higher phosphate release-to-uptake ratio and, respectively, in a much higher phosphorus removal efficiency (93.8%) than POSBR2 (84.7%) and POSBR3 (77.3%) within 30 days of operation. This indicated that the phosphorus removal efficiency decreased the higher the starch concentration was. It was also found that POSBR1 produced more polyhydroxyalkanoates (PHAs) than the other reactors. Based on the effect of the carbon source on the PHA concentration and consumption, the conditions of POSBR1 were favourable for the growth of polyphosphate-accumulating organisms and therefore, beneficial for the biological phosphorus removal process.

  14. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    National Research Council Canada - National Science Library

    Dehjourian, Mehdi; Rahgoshay, Mohammad; Sayareh, Reza; Jahanfarnia, Gholamreza; Shirani, Amir Saied

    2016-01-01

    The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code...

  15. A high-pressure plug flow reactor for combustion chemistry investigations

    Science.gov (United States)

    Lu, Zhewen; Cochet, Julien; Leplat, Nicolas; Yang, Yi; Brear, Michael J.

    2017-10-01

    A plug flow reactor (PFR) is built for investigating the oxidation chemistry of fuels at up to 50 bar and 1000 K. These conditions include those corresponding to the low temperature combustion (i.e. the autoignition) that commonly occurs in internal combustion engines. Turbulent flow that approximates ideal, plug flow conditions is established in a quartz tube reactor. The reacting mixture is highly diluted by excess air to reduce the reaction rates for kinetic investigations. A novel mixer design is used to achieve fast mixing of the preheated air and fuel vapour at the reactor entrance, reducing the issue of reaction initialization in kinetic modelling. A water-cooled probe moves along the reactor extracting gases for further analysis. Measurement of the sampled gas temperature uses an extended form of a three-thermocouple method that corrects for radiative heat losses from the thermocouples to the enclosed PFR environment. Investigation of the PFR’s operation is first conducted using non-reacting flows, and then with isooctane oxidation at 900 K and 10 bar. Mixing of the non-reacting temperature and species fields is shown to be rapid. The measured fuel consumption and CO formation are then closely reproduced by kinetic modelling using an extensively validated iso-octane mechanism from the literature and the corrected gas temperature. Together, these results demonstrate the PFR’s utility for chemical kinetic investigations.

  16. 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)

  17. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  18. Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

    Energy Technology Data Exchange (ETDEWEB)

    Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

    2017-03-15

    Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

  19. Influence of acidogenic headspace pressure on methane production under schematic of diversion of acidogenic off-gas to methanogenic reactor.

    Science.gov (United States)

    Yan, Bing Hua; Selvam, Ammaiyappan; Wong, Jonathan W C

    2017-12-01

    This study investigated the effects of 12.6psi (T1), 6.3psi (T2), 3.3psi (T3) and ambient (T4) headspace pressure on the metabolic pathways in the acidogenic leach bed reactor (LBR) and overall methane recovery during two-phase anaerobic digestion of food waste. Diversion of biogas from LBR enhanced COD and soluble product generation in T2, T3 and T4 whereas, high pressure (T1) resulted in comparatively higher lactate production and low protein degradation. A pressure of 3-6psi (T2 and T3) improved the production of COD by ∼22-36%, soluble products by ∼9-43%, volatile solid reduction by ∼14-19%, and CH4 production by ∼10-31% compared with control. Besides, ∼3-6psi headspace pressure positively influenced the composition of soluble products resulting in enhanced methane recovery adding advantage to the two-phase system. A headspace pressure of ∼3-6psi is recommended to enhance the hydrolysis-acidogenesis; however, the actual hydrogen concentration should be considered. Copyright © 2017. Published by Elsevier Ltd.

  20. 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.

  1. Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

    Science.gov (United States)

    Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

    2018-01-01

    China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

  2. General installation of pressurized water reactors; Installation generale des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, M. [Electricite de France (EDF), 75 - Paris (France)

    2002-07-01

    This article describes the criteria considered for the installation of PWR reactors according to the most recent French standard: the N4 - 1450 MWe plant series which benefits of the experience feedback of the previous 900 and 1300 MWe series: 1 - functional organization: nuclear island, classical island; 2 - general principles of installation: safety, functional and operation requirements, grouping of equipments inside buildings, criteria of equipments separation; 3 - location of the main equipments, plot plan: specific site constraints, sitting of facilities, leveling; 4 - general installation of buildings: reactor building, fuel building, electrical and auxiliaries building, nuclear auxiliaries building, machines room, pumping station, standby generators building, water storage tank and pool building, effluents processing building; 5 - the EPR european concept: 4 sub-series concept, prevention of serious accidents, annual collective dose. (J.S.)

  3. 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.)

  4. 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.

  5. Solution to the incompatibility between reactor protection logic and turbine shot logic. Scram by high pressure; Solucion a la incompatibilidad entre logica de proteccion de reactor y logica de disparo de turbina. SCRAM por alta presion

    Energy Technology Data Exchange (ETDEWEB)

    Ramos Q, R.; Santiago F, C.; Gonzalez P, G., E-mail: ruben.ramos01@cfe.gob.mx [Comision Federal de Electricidad, Central Nuclear Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    The nuclear power plant of Laguna Verde carried out the Modernization and Increase of Extended Power Project in its two Units (2005-2011). This modernization included to the electro-hydraulic control system of the main turbine, replacing an ana logical system by one digital (Digital Electro-hydraulic Control - DEHC) whose functions are of controlling the reactor pressure in the different operation ways as wells as of controlling the velocity and load of the main turbine. Also, it has protections that are related with diverse plant systems, as the Reactor Protection Systems (RPS). During the tests stage was realized a programmed load rejection, which Reactor Scram should cause when being presented the shot of main turbine. However, the logic of the RPS was inhibited due to the quick response of the new control DEHC, propitiating a condition of non prospective plant and, in consequence, the Reactor Scram happened for another protection of the RPS. (author)

  6. The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Ruscak, M.; Chvatal, P.; Karnik, D.

    1997-04-01

    One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

  7. Steady-state CFD simulations of an EPR™ reactor pressure vessel: A validation study based on the JULIETTE experiments

    Energy Technology Data Exchange (ETDEWEB)

    Puragliesi, R., E-mail: riccardo.puragliesi@psi.ch [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland); Zhou, L. [Science and Technology on Reactor System Design Technology Laboratory, NPIC, Chengdu (China); Zerkak, O.; Pautz, A. [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland)

    2016-04-15

    Highlights: • CFD validation of k–ε (RANS model of EPR RPV. • Flat inlet velocity profile is not sufficient to correctly predict the pressure drops. • Swirl is responsible for asymmetric loads at the core barrel. • Parametric study to the turbulent Schmidt number for better predictions of passive-scalar transport. • The optimal turbulent Schmidt number was found to be one order of magnitude smaller than the standard value. - Abstract: Validating computational fluid dynamics (CFD) models against experimental measurements is a fundamental step towards a broader acceptance of CFD as a tool for reactor safety analysis when best-estimate one-dimensional thermal-hydraulic codes present strong modelling limitations. In the present paper numerical results of steady-state RANS analyses are compared to pressure, volumetric flow rate and concentration distribution measurements in different locations of an Areva EPR™ reactor pressure vessel (RPV) mock-up named JULIETTE. Several flow configurations are considered: Three different total volumetric flow rates, cold leg velocity field with or without swirl, three or four reactor coolant pumps functioning. Investigations on the influence of two types of inlet boundary profiles (i.e. flat or 1/7th power-law) and the turbulent Schmidt number have shown that the first affects sensibly the pressure loads at the core barrel whereas the latter parameter strongly affects the transport and the mixing of the tracer (passive scalar) and consequently its distribution at the core inlet. Furthermore, the introduction of an integral parameter as the swirl number has helped to decrease the large epistemic uncertainty associated with the swirling device. The swirl is found to be the cause of asymmetric loads on the walls of the core barrel and also asymmetries are enhanced for the tracer concentration distribution at the core inlet. The k–ϵ CFD model developed with the commercial code STAR-CCM+ proves to be able to predict

  8. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  9. A simple kinetic analysis of syngas during steam hydrogasification of biomass using a novel inverted batch reactor with instant high pressure feeding.

    Science.gov (United States)

    Fan, Xin; Liu, Zhongzhe; Norbeck, Joseph M; Park, Chan S

    2016-01-01

    A newly designed inverted batch reactor equipped with a pressure-driven feeding system was built for investigating the kinetics of syngas during the steam hydrogasification (SHR) of biomass. The system could instantly load the feedstock into the reactor at high temperature and pressure, which simulated the way to transport the feedstock into a hot and pressurized gasifier. Experiments were conducted from 600°C to 700°C. The inverted reactor showed very high heating rate by enhancing the carbon conversion and syngas production. The kinetic study showed that the rates of CH4, CO and CO2 formation during SHR were increased when the gasification temperature went up. SHR had comparatively lower activation energy for CH4 production. The activation energies of CH4, CO and CO2 during SHR were 42.8, 51.8 and 14kJ/mol, respectively. Copyright © 2015 Elsevier Ltd. All rights reserved.

  10. Temperature uniformity mapping in a high pressure high temperature reactor using a temperature sensitive indicator

    NARCIS (Netherlands)

    Grauwet, T.; Plancken, van der I.; Vervoort, L.; Matser, A.M.; Hendrickx, M.; Loey, van A.

    2011-01-01

    Recently, the first prototype ovomucoid-based pressure–temperature–time indicator (pTTI) for high pressure high temperature (HPHT) processing was described. However, for temperature uniformity mapping of high pressure (HP) vessels under HPHT sterilization conditions, this prototype needs to be

  11. Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    of the mixture was varied from oxidizing to reducing conditions. Moreover, a series of experiments in an oxygen atmosphere instead of a nitrogen atmosphere has been done. A reaction mechanism based on a recent work by Burke et al. has been developed. In addition to modeling of the present experiments......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

  12. Understanding the role of defect production in radiation embrittlement of reactor pressure vessels.

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, D. E.

    1999-08-04

    Comparative experiments using high energy (10 MeV) electrons and test reactor neutrons have been undertaken to understand the role that primary damage state has on hardening (embrittlement) induced by irradiation at 300 C. Electrons produce displacement damage primarily by low energy atomic recoils, while fast neutrons produce displacements from considerably higher energy recoils. Comparison of changes resulting from neutron irradiation, in which nascent point defect clusters can form in dense cascades, with electron irradiation, where cascade formation is minimized, can provide insight into the role that the in-cascade point defect clusters have on the mechanisms of embrittlement. Tensile property changes induced by 10 MeV electrons or test reactor neutron irradiations of unalloyed iron and an Fe-0.9 wt.% Cu-1.0 wt.% Mn alloy were examined in the damage range of 9.0 x 10{sup {minus}5} dpa to 1.5 x 10{sup {minus}2} dpa. The results to date showed the ternary alloy experienced substantially greater embrittlement in both the electron and neutron irradiated samples relative to unalloyed iron. Surprisingly, despite their disparate nature of defect production, similar embrittlement trends with increasing radiation damage were observed for electrons and neutrons in both the ternary and unalloyed iron.

  13. Atmospheric Pressure Non-Thermal Plasma Activation of CO2in a Packed-Bed Dielectric Barrier Discharge Reactor.

    Science.gov (United States)

    Mei, Danhua; Tu, Xin

    2017-11-17

    Direct conversion of CO 2 into CO and O 2 is performed in a packed-bed dielectric barrier discharge (DBD) non-thermal plasma reactor at low temperatures and atmospheric pressure. The maximum CO 2 conversion of 22.6 % is achieved when BaTiO 3 pellets are fully packed into the discharge gap. The introduction of γ-Al 2 O 3 or 10 wt % Ni/γ-Al 2 O 3 catalyst into the BaTiO 3 packed DBD reactor increases both CO 2 conversion and energy efficiency of the plasma process. Packing γ-Al 2 O 3 or 10 wt % Ni/γ-Al 2 O 3 upstream of the BaTiO 3 bed shows higher CO 2 conversion and energy efficiency compared with that of mid- or downstream packing modes because the reverse reaction of CO 2 conversion-the recombination of CO and O to form CO 2 -is more likely to occur in mid- and downstream modes. Compared with the γ-Al 2 O 3 support, the coupling of the DBD with the Ni catalyst shows a higher CO 2 conversion, which can be attributed to the presence of Ni active species on the catalyst surface. The argon plasma treatment of the reacted Ni catalyst provides extra evidence to confirm the role of Ni active species in the conversion of CO 2 . © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Soo-Yong Park

    2015-10-01

    Full Text Available Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  15. Methods and performance of the three-dimensional pressurized water reactor core dynamics SIMTRAN on-line code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J.M.; Ahnert, C.; Cabellos, O. [Polytechnic Univ., Madrid (Spain)

    1996-09-01

    New reactor physics and computation methods have been developed in the three-dimensional pressurized water reactor (PWR) core dynamics SIMTRAN code for on-line surveillance and prediction. The accuracy of the coupled neutronic thermal-hydraulic solution is improved, and its scope is extended to provide, mainly, the calculation of the fission reaction rates at the in-core mini detectors, the responses at the ex-core detectors, and the in-vessel coolant flow and temperature distributions. The functional capabilities implemented in the on-line SIMTRAN code include on-line surveillance, in-core-ex-core calibration, evaluation of peak power factors and thermal margins, nominal cycle follow, prediction of maneuvers, and diagnosis of fast transients and oscillations. The new code has been operating on-line at the Vandellos-II PWR unit in Spain since the startup of its cycle 7 in mid-June 1994, including the machine-man interfaces for on-line acquisition of measured data and interactive graphical utilization. The agreement of the simulations with the measurements, along the full cycle 7 and the first months of cycle 8 operation, is well within the accuracy requirements. The performance and usefulness for operational support shown during the demo and routine use phases have proved that the on-line SIMTRAN code has the qualities for the accurate, reliable, comprehensive, and user-friendly on-line core surveillance and prediction.

  16. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    Directory of Open Access Journals (Sweden)

    Jung Suk Kim

    2015-12-01

    Full Text Available The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233U, 242Pu, 150Nd, and 133Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

  17. 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.

  18. Analysis of quench-vent pressures for present design of ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coils

    Energy Technology Data Exchange (ETDEWEB)

    Slack, D.S.

    1989-09-20

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three- dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 1 fig., 3 tabs.

  19. A Novel Fractional-Order PID Controller for Integrated Pressurized Water Reactor Based on Wavelet Kernel Neural Network Algorithm

    Directory of Open Access Journals (Sweden)

    Yu-xin Zhao

    2014-01-01

    Full Text Available This paper presents a novel wavelet kernel neural network (WKNN with wavelet kernel function. It is applicable in online learning with adaptive parameters and is applied on parameters tuning of fractional-order PID (FOPID controller, which could handle time delay problem of the complex control system. Combining the wavelet function and the kernel function, the wavelet kernel function is adopted and validated the availability for neural network. Compared to the conservative wavelet neural network, the most innovative character of the WKNN is its rapid convergence and high precision in parameters updating process. Furthermore, the integrated pressurized water reactor (IPWR system is established by RELAP5, and a novel control strategy combining WKNN and fuzzy logic rule is proposed for shortening controlling time and utilizing the experiential knowledge sufficiently. Finally, experiment results verify that the control strategy and controller proposed have the practicability and reliability in actual complicated system.

  20. Corrosion Behavior of Carbon Steel Coated with Octadecylamine in the Secondary Circuit of a Pressurized Water Reactor

    Science.gov (United States)

    Jäppinen, Essi; Ikäläinen, Tiina; Järvimäki, Sari; Saario, Timo; Sipilä, Konsta; Bojinov, Martin

    2017-10-01

    Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

  1. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used.

  2. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    Science.gov (United States)

    Sun, Mingyue; Hao, Luhan; Li, Shijian; Li, Dianzhong; Li, Yiyi

    2011-11-01

    Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  3. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    Energy Technology Data Exchange (ETDEWEB)

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  4. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  5. Hydrodynamics in a pressurized cocurrent gas-liquid trickle bed reactor

    NARCIS (Netherlands)

    Wammes, Wino J.A.; Westerterp, K.R.

    1991-01-01

    The influence of gas density on total external liquid hold-up, pressure drop and gas-liquid interfacial area, under trickle-flow conditions, and the transition to pulse flow have been investigated with nitrogen or helium as the gas phase up to 7.5 MPa. It is concluded that the hydrodynamics depends

  6. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  7. Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States); Shayer, Z., E-mail: zshayer@mines.edu [Department of Physics, Colorado School of Mines, 1500 Illinois St., Golden, CO 80401 (United States)

    2017-03-15

    Highlights: • We evaluate transmutation fuels for plutonium and minor actinide destruction in LWRs. • We model a modified AP1000 fuel assembly in SCALE6.1. • We evaluate spectral shift absorber coatings to improve transmutation performance. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO{sub 2}, Pu{sub 3}Si{sub 2}, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO{sub 2}) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5 wt% PuO{sub 2}), Pu{sub 0.31}ZrH{sub 1.6}Th{sub 1.08}, and PuZrO{sub 2}MgO (8 wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B{sub 4}C, CdO, Dy{sub 2}O{sub 3}, Er{sub 2}O{sub 3}, Eu{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, HfO{sub 2}, In{sub 2}O{sub 3}, Lu{sub 2}O{sub 3}, Sm{sub 2}O{sub 3}, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO{sub 2}MgO (8 wt% Pu) target

  8. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J. [Nuclear Res. Inst., Rez (Czech Republic)

    1998-11-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  9. Addressing the scaling issue with Cathare 2 simulation of VVER 1000 transient scenario

    Energy Technology Data Exchange (ETDEWEB)

    Dino Araneo; Alessandro Del Nevo; Francesco D' Auria; Giorgio Galassi [DIMNP Universita di Pisa, Via Diotisalvi, 2, 56122 Pisa (Italy)

    2005-07-01

    Full text of publication follows: Tests performed at scaled facilities play an important role in the assessment of safety of Nuclear Power Plant (NPP). The results obtained by the tests performed in the facilities can be used to qualifies the NPP nodalization. Starting from the same initial and boundary conditions of the experimental tests performed at the facility the full plant nodalization must reproduce the same phenomena with the same timing. This is indicated as 'Kv scaled calculation'. A good agreement between the results obtained in the calculation and the experimental tests allows to say that the plant nodalization is able to reproduce the behaviour of the plant in transient scenarios. This paper deals with the scaling issue concerning a Cathare2 simulation of a VVER 1000 transient scenario. The PSB-VVER facility is built in 1998 at Electrogorsk Research and Engineering Centre. It is a facility with a scaling factor of 1/300 for the volume of the referred NPP (VVER-1000). In order to evaluate the nodalization performance the qualification procedure developed at the DIMNP of Pisa University (UNIPI) has been applied. This procedure foresees two levels of qualification: a 'steady state' level and an 'on transient' level. After the steady state results of the nodalization has been checked, the 'on transient' qualification check is performed adopting the PSB-VVER 11% equivalent break in Upper Plenum. This test includes the actuation of the HPIS injecting only in the loop 4 and the availability of the hydro accumulators. (authors)

  10. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, J.H.; McCauley, E.W.

    1977-12-22

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a /sup 1///sub 5/-scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the /sup 1///sub 5/-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor.

  11. A vented low pressure containment strategy for the Modular High Temperature Gas-Cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Dilling, D. [Bechtel National, Inc., San Francisco, CA (United States); Dunn, T.D.; Silady, F.A. [General Atomics, San Diego, CA (United States)

    1994-04-01

    This paper presents the response of the 450 MW(t) MHTGR with a steam turbine power conversion system to expected and hypothetical accident source term assumptions. A range of vented low pressure containment (VLPC) strategies was considered that would enhance the retention of radionuclides. This study was prepared to review the technical merits of VLPC options in response to an NRC request during preapplication review of the steam cycle MMGR. The study found that, even under arbitrary hypothetical assumptions regarding significantly lower than expected fuel performance, vented low pressure containment options can effectively reduce accident doses. The reference design with a VLPC meets the 10CFR100 and prompt fatality doses even with lower than expected fuel performance. Alternative VLPC designs were studied which could be used to augment the current design to provide additional margin.

  12. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  13. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  14. 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

  15. Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    Energy Technology Data Exchange (ETDEWEB)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  16. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  17. Reactor Design for CO2 Photo-Hydrogenation toward Solar Fuels under Ambient Temperature and Pressure

    Directory of Open Access Journals (Sweden)

    Chun-Ying Chen

    2017-02-01

    Full Text Available Photo-hydrogenation of carbon dioxide (CO2 is a green and promising technology and has received much attention recently. This technique could convert solar energy under ambient temperature and pressure into desirable and sustainable solar fuels, such as methanol (CH3OH, methane (CH4, and formic acid (HCOOH. It is worthwhile to mention that this direction can not only potentially depress atmospheric CO2, but also weaken dependence on fossil fuel. Herein, 1 wt % Pt/CuAlGaO4 photocatalyst was successfully synthesized and fully characterized by ultraviolet-visible light (UV-vis spectroscopy, X-ray diffraction (XRD, Field emission scanning electron microscopy using energy dispersive spectroscopy analysis (FE-SEM/EDS, transmission electron microscopy (TEM, X-ray photoelectron spectroscopy (XPS, and Brunauer-Emmett-Teller (BET, respectively. Three kinds of experimental photo-hydrogenation of CO2 in the gas phase, liquid phase, and gas-liquid phase, correspondingly, were conducted under different H2 partial pressures. The remarkable result has been observed in the gas-liquid phase. Additionally, increasing the partial pressure of H2 would enhance the yield of product. However, when an extra amount of H2 is supplied, it might compete with CO2 for occupying the active sites, resulting in a negative effect on CO2 photo-hydrogenation. For liquid and gas-liquid phases, CH3OH is the major product. Maximum total hydrocarbons 8.302 µmol·g−1 is achieved in the gas-liquid phase.

  18. Development of an improved methodology for the determination of the neutron load of the pressure vessel steel of WWER-1000 type reactors. Final report; Entwicklung einer fortgeschrittenen Methodik zur Bestimmung der Neutronenbelastung des Druckbehaeltermaterials vom Reaktor des Typs WWER-1000. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Barz, H.U.; Boehmer, B.; Konheiser, J.; Stephan, I.

    1998-05-01

    The investigations, reported on here, are aimed at the theoretical and experimental determination of the neutron exposure of the VVER-1000 reactors Balakovo-3 and Rovno-3. The overall approach, partially developed and at least improved in the frame of this project, comprises the pure calculation part, the gamma spectrometric analysis of neutron activation detectors and the comparison of experimental and theoretical results using the spectrum adjustment procedure. The approach is not restricted to neutron embrittlement but can be applied to neutron fluence problems in general. (orig./GL) [Deutsch] In diesem Projekt wurden fuer die WWER-1000 Reaktoren Balakovo-3 und Rovno-3 die Parameter der Neutronenbelastung experimentell und theoretisch bestimmt. Der vorliegende Bericht beschreibt das methodische Vorgehen, welches aus dem reinen Berechnungsteil, der gammaspektrometrischen Analyse der Ativierungsdetektoren und dem Vergleich der gemessenen und berechneten Werte einschliesslich der Spektrumsjustierung besteht. Dieses Instrumentarium, welches allgemein bei der Bestimmung der Neutronenfluenz anwendbar ist, wurde im Projektzeitraum weiter verbessert. (orig./GL)

  19. The use of cyclone filters to reduce the foreign matter insertion into the reactor pressure vessel; Einsatz von Zyklonabscheidern zur Verringerung eines Fremdkoerpereintrags in den Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Widmann, W. [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2010-05-15

    Fuel element damage can induce a significant disturbance of the normal NPP operation. Foreign matter insertion into the reactor pressure vessel by the feed water pipes is supposed to be the main cause for fuel element damage in boiling water reactors. Vattenfall AB and Westinghouse Electric Sweden AB have developed a cyclone filter for nuclear facilities. The operational experience of cyclone filters in the feedwater line have reduced the fuel element damages significantly. Westinghouse Electric Germany is performing a feasibility study for the implementation of cyclone filter in the NPP Kruemmel.

  20. Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels; Evaluacion computacional del efecto de la perdida de constriccion en la tenacidad de fractura de la vasija de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Serrano Garcia, M.

    2007-07-01

    The Master Curve approach is included on the ASME Code through some Code Cases to assess the reactor pressure vessel integrity. However, the margin definition to be added is not defined as is the margin to be added when the Master Curve reference temperature T{sub 0} is obtained by testing pre-cracked Charpy specimens. The reason is that the T{sub 0} value obtained with this specimen geometry is less conservative than the value obtained by testing compact tension specimens possible due to a loss of constraint. The two parameter fracture mechanics, considered as an extension of the classical fracture mechanics, coupled to a micromechanical fracture models is a valuable tool to assess the effect of constraint loss on fracture toughness. The definition of a parameter able to connect the fracture toughens value to the constraint level on the crack tip will allow to quantify margin to be added to the T{sub 0} value when this value is obtained testing the pre-cracked Charpy specimens included in the surveillance capsule of the reactor pressure vessel. The Nuclear Regulatory Commission (NRC) define on the To value obtained by testing compact tension specimens and ben specimens (as pre-cracked Charpy are) bias. the NRC do not approved any of the direct applications of the Master Curve the reactor pressure vessel integrity assessment until this bias will be quantified in a reliable way. the inclusion of the bias on the integrity assessment is done through a margin to be added. In this thesis the bias is demonstrated an quantified empirical and numerically and a generic value is suggested for reactor pressure vessel materials, so that it can be used as a margin to be added to the T{sub 0} value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref.

  1. A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A., E-mail: aaron@epiney.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Service d' Etudes des Systemes Innovants, F-13108 Saint Paul Lez Durance (France); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer An analytical model predicting Brayton cycle off-design steady states, is developed. Black-Right-Pointing-Pointer The model is used to design an autonomous decay heat removal system for the GFR. Black-Right-Pointing-Pointer Predictions of the analytical model are verified using CATHARE. Black-Right-Pointing-Pointer CATHARE code is used to simulate a set of GFR safety depressurization transients using this device. Black-Right-Pointing-Pointer Convenient turbo-machine designs exist for the targeted autonomous decay heat removal for a wide pressure range. - Abstract: This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping model', is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the

  2. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  3. Methane conversion using a dielectric barrier discharge reactor at atmospheric pressure for hydrogen production

    Science.gov (United States)

    Khadir, N.; Khodja, K.; Belasri, A.

    2017-09-01

    In the present paper, we carried out a theoretical study of dielectric barrier discharge (DBD) filled with pure methane gas. The homogeneous discharge model used in this work includes a plasma chemistry unit, an electrical circuit, and the Boltzmann equation. The model was applied to the case of a sinusoidal voltage at a period frequency of 50 kHz and under a gas pressure of 600 Torr. We investigated the temporal variation of electrical and kinetic discharge parameters such as plasma and dielectric voltages, the discharge current density, electric field, deposited power density, and the species concentration. We also checked the physical model validity by comparing its results with experimental work. According to the results discussed herein, the dielectric capacitance is the parameter that has the greatest effect on the methane conversion and H2/CH4 ratio. This work enriches the knowledge for the improvement of DBD for CH4 conversion and hydrogen production.

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

    Science.gov (United States)

    Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Matsukawa, Y.; Nishiyama, Y.; Onizawa, K.

    2012-06-01

    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 δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °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 δ-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 γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

  5. Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

    Energy Technology Data Exchange (ETDEWEB)

    Lowendahl, B. [OKG Aktiebolag, Figeholm (Sweden)

    1995-03-01

    In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large project was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.

  6. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  7. Modeling of the Radiation Doses during Dismantling of RBMK-1500 Reactor Pressurized Tanks from Emergency Core Cooling System

    Directory of Open Access Journals (Sweden)

    A. Simonis

    2013-01-01

    Full Text Available Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. One of them is personnel radiation safety during the performance of dismantling activities. In this paper, modeling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT of RBMK-1500 reactor are presented. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The effective doses to the workers have been modeled for different strategies of ECCS PT dismantling. In order to select the optimal personnel radiation safety, the modeling has been performed by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium. The impacts of dismantling tools, shielding types, and extract ventilation flow rate on effective doses during the dismantling of ECCS PT have been analyzed. The total effective personnel doses have been obtained by summarizing the effective personnel doses from various sources of exposure, that is, direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises. The uncertainty of the collective doses is also presented in this paper.

  8. Simulation Analysis of Sludge Disposal and Volatile Fatty Acids Production from Gravity Pressure Reactor via Wet Air Oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gwon Woo [Biomass and Waste Energy Laboratory, KIER, Daejeon (Korea, Republic of); Seo, Tae Wan; Lee, Hong-Cheol; Hwang, In-Ju [Environmental and Plant Engineering Research Institute, KICT, Goyang (Korea, Republic of)

    2016-04-15

    Efficacious wastewater treatment is essential for increasing sewage sludge volume and implementing strict environmental regulations. The operation cost of sludge treatment amounts up to 50% of the total costs for wastewater treatment plants, therefore, an economical sludge destruction method is crucially needed. Amid several destruction methods, wet air oxidation (WAO) can efficiently treat wastewater containing organic pollutants. It can be used not only for sludge destruction but also for useful by-product production. Volatile fatty acids (VFAs), one of many byproducts, is considered to be an important precursor of biofuel and chemical materials. Its high reaction condition has instituted the study of gravity pressure reactor (GPR) for an economical process of WAO to reduce operation cost. Simulation of subcritical condition was conducted using Aspen Plus with predictive Soave-Redlich-Kwong (PSRK) equation of state. Conjointly, simulation analysis for GPR depth, oxidizer type, sludge flow rate and oxidizer injection position was carried out. At GPR depth of 1000m and flow rate of 2 ton/h, the conversion and yield of VFAs were 92.02% and 0.17g/g, respectively.

  9. 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)

  10. Effect of Heavy Ion Irradiation Dosage on the Hardness of SA508-IV Reactor Pressure Vessel Steel

    Directory of Open Access Journals (Sweden)

    Xue Bai

    2017-01-01

    Full Text Available Specimens of the SA508-IV reactor pressure vessel (RPV steel, containing 3.26 wt. % Ni and just 0.041 wt. % Cu, were irradiated at 290 °C to different displacement per atom (dpa with 3.5 MeV Fe ions (Fe2+. Microstructure observation and nano-indentation hardness measurements were carried out. The Continuous Stiffness Measurement (CSM of nano-indentation was used to obtain the indentation depth profile of nano-hardness. The curves showed a maximum nano-hardness and a plateau damage near the surface of the irradiated samples, attributed to different hardening mechanisms. The Nix-Gao model was employed to analyze the nano-indentation test results. It was found that the curves of nano-hardness versus the reciprocal of indentation depth are bilinear. The nano-hardness value corresponding to the inflection point of the bilinear curve may be used as a parameter to describe the ion irradiation effect. The obvious entanglement of the dislocations was observed in the 30 dpa sample. The maximum nano-hardness values show a good linear relationship with the square root of the dpa.

  11. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 10{sup 19} n cm{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 13} n cm{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  12. Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Suzuki, M. [Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan)

    2014-09-15

    The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400 °C for periods of 100–10,000 h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100 h, and gradually for aging durations greater than 1000 h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000 h, had the highest number density after aging for 2000 h, and coarsened after aging for 10,000 h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000 h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition.

  13. A RETRAN model of the Calvert Cliffs-1 pressurized water reactor for assessing the safety implications of control systems

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J P.A.; Smith, O L

    1987-03-01

    The failure mode and effects analysis of Calvert Cliffs-1 identified sequences of events judged sufficiently complex to merit further analysis in detailed dynamic simulations. This report describes the RETRAN model developed for this purpose and the results obtained. The mathematical tool was RETRAN2/Mod3, the latest version of a widely used and extensively validated thermal-hydraulics production code obtained by license agreement with the developer, Electric Power Research Institute, and installed on the ORNL BM-3033 computers. RETRAN2 is based on a first-principles methodology that treats two-phase flow with slip. Thermal equilibrium of phases is assumed except in the pressurizer, where non-equilibrium processes are important and special methodology is used. Heat transfer in solids is obtained from the conventional conduction equation. Point or 1-D kinetics is available for the reactor core. The fundamental methodology is supplemented with a broad list of process submodels that calculate heat transfer coefficients, fluid and metal state properties, choked flow, form and wall friction losses, and other parameters. Also supplied are component submodels for various types of valves and pumps, the latter of which incorporate four-quadrant characteristics for components in which two-phase or reverse flow may be expected, and head versus flow curves for others. Extensive input allows the code to be highly particularized to a specific plant. The major investment in time and manpower occurs in setting up the base case; changes are comparatively easy to implement.

  14. Validation of the pressurized water reactor core analysis system SEANAP-86 with measurements in tests and operation

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Aragones, J.M.; Crespo, A.; Labay, A.; Leon, J.R.; Alvarez, A.I.

    1988-11-01

    The advances achieved during 1986 in the development and validation of the Spanish pressurized water reactor (PWR) core analysis system SEANAP-86 are presented and discussed. The advances were achieved through the cooperation of two research institutes (DENIM and JENCIEMAT) and two nuclear power plants (Almaraz and Asco). The main features and advances of the core analysis system are summarized. The system is intended to provide the utilities with the in-house capability to analyze the core design, nuclear tests, and operation along reload cycles, as technical support for licensing and operation. The validation is done by an extensive comparison with the measurements available from the nuclear tests and in-core operation followup of four 900-MW(electric)PWR units (Almaraz-1 and -2, Asco-1 and -2) along their first cycles and several reload cycles, as well as with the available design data. Good agreement is obtained for the broad sets of parameters and cycles analyzed, which qualifies the SEANAP-86 system as an effective core analysis support tool and encourages further developments, which are presently under way.

  15. Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

    Energy Technology Data Exchange (ETDEWEB)

    Han, Guangdong [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, Shanghai, 200072 (China); Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China)

    2015-12-15

    The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen. - Highlights: • Duplex oxide films on ground and mechanically polished specimens. • Compact oxide on electropolished specimen after 120 h immersion. • Large spinel outer layer rich in Fe and fine spinel inner layer rich in Cr. • Electropolishing improved oxidation resistance especially at the early stages. • Inhomogeneous Cr-rich inner layer with granular areas affected by surface treatment.

  16. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  17. STC Germany/Russia. Fluence calculations of surveillance specimens of the VVER-440. Final report; WTZ Russland. Fluenzberechnungen fuer Voreilproben beim WWER-440. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Konheiser, J.; Grahn, A.

    2014-07-01

    Reactor pressure vessels (RPV) are non-restorable equipment and their lifetime may restrict the nuclear power plant-life as a whole. Surveillance specimen programs for RPV materials are among the most important measures of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV residual lifetime. In addition to the chemical composition of the RPV steel, the radiation parameters (neutron and gamma fluences and spectra) have the most important impact on the RPV embrittlement characteristics. In this work, different geometric positions which have influence on the radiation conditions of the samples are investigated. Thus, the uncertainties can be determined in the fluence values of surveillance specimens. The fluence calculations were carried out by the codes TRAMO and DORT. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP, Unit 3 (VVER-440/213), which provide the basis for validation of calculated neutron fluences. The main neutron-activation monitoring reactions were {sup 54}Fe(n,p){sup 54}Mn and {sup 58}Ni(n,p){sup 58}Co. The activity measurements were carried out by ''Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS). Good agreement between the deterministic and stochastic calculation results as well as between the calculations and the ex-vessel measurements was found. The average difference between measured and calculated values is 5%. The influence of the channels for surveillance specimens and the shielding effect of a baffle rib on the monitors and on the Monte-Carlo calculated results was studied. For the surveillance specimens in the maximum of the flux, an average flux of around 2.45 * 10{sup 12} neutrons/cm{sup 2} was calculated for the neutron flux E> 0.5 MeV. The differences in the surveillance specimens could be up to 20% depending on the direction to the core. Discrepancies up to 10% can be caused by the change of the position of the

  18. 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.

  19. 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.

  20. 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.

  1. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  2. 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

  3. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    Science.gov (United States)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

    Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

  10. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    Science.gov (United States)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    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.

  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 (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

  13. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  14. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  15. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  16. Assessing the impact of electrolyte conductivity and viscosity on the reactor cost and pressure drop of redox-active polymer flow batteries

    Science.gov (United States)

    Iyer, Vinay A.; Schuh, Jonathon K.; Montoto, Elena C.; Pavan Nemani, V.; Qian, Shaoyi; Nagarjuna, Gavvalapalli; Rodríguez-López, Joaquín; Ewoldt, Randy H.; Smith, Kyle C.

    2017-09-01

    Redox-active small molecules, used traditionally in redox flow batteries (RFBs), are susceptible to crossover and require expensive ion exchange membranes (IEMs) to achieve long lifetimes. Redox-active polymer (RAP) solutions show promise as candidate electrolytes to mitigate crossover through size-exclusion, enabling the use of porous separators instead of IEMs. Here, poly(vinylbenzyl ethyl viologen) is studied as a surrogate RAP for RFBs. For oxidized RAPs, ionic conductivity varies weakly between 1.6 and 2.1 S m-1 for RAP concentrations of 0.13-1.27 mol kg-1 (monomeric repeat unit per kg solvent) and 0.32 mol kg-1 LiBF4 with a minor increase upon reduction. In contrast, viscosity varies between 1.8 and 184.0 mPa s over the same concentration range with weakly shear-thinning rheology independent of oxidation state. Techno-economic analysis is used to quantify reactor cost as a function of electrolyte transport properties for RAP concentrations of 0.13-1.27 mol kg-1, assuming a hypothetical 3V cell and facile kinetics. Among these concentrations, reactor cost is minimized over a current density range of 600-1000 A m-2 with minimum reactor cost between 11-17 per kWh, and pumping pressures below 10 kPa. The predicted low reactor cost of RAP RFBs is enabled by sustained ionic mobility in spite of the high viscosity of concentrated RAP solutions.

  17. Characterization of a novel micro-pressure swirl reactor for removal of chemical oxygen demand and total nitrogen from domestic wastewater at low temperature.

    Science.gov (United States)

    Ren, Qingkai; Yu, Yang; Zhu, Suiyi; Bian, Dejun; Huo, Mingxin; Zhou, Dandan; Huo, Hongliang

    2017-06-01

     A novel micro-pressure swirl reactor (MPSR) was designed and applied to treat domestic wastewater at low temperature by acclimating microbial biomass with steadily decreasing temperature from 15 to 3 °C. Chemical oxygen demand (COD) was constantly removed by 85% and maintained below 50 mg L-1 in the effluent during the process. When the air flow was controlled at 0.2 m3 h-1, a swirl circulation was formed in the reactor, which created a dissolved oxygen (DO) gradient with a low DO zone in the center and a high DO zone in the periphery for denitrification and nitrification. 81% of total nitrogen was removed by this reactor, in which ammonium was reduced by over 90%. However, denitrification was less effective because of the presence of low levels of oxygen. The progressively decreasing temperature favored acclimation of psychrophilic bacteria in the reactor, which replaced mesophilic bacteria in the process of treatment.

  18. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  19. Optimization of the Manufacturing Process of Zr-2.5Nb Pressure Tubes for CANDU Reactors for Extending Their Design Life to Over 30 years

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    Zr-2.5Nb pressure tubes are the most critical components that determine the design life of CANADU (CAnadian Natural Uranium) reactors. The initial design target for the Zr-2.5Nb pressure tubes is to suppress the diametral creep through a texture control which may trade off the other performances that can be overcome by introducing a change in the components design. To this end, they are made by the extrusion process at high temperatures to have a circumferential texture with most of the basal poles oriented towards their circumferential direction. However, this circumferential texture causes them to be very susceptible to delayed hydride cracking (DHC) and to have a higher axial elongation. Against the initial design target, their costly refurbishments are planned in several commercial CANDU reactors before their design life of 30 years, due to the unexpectedly faster creep rate and axial elongation. This fact casts a question over the validity of the design philosophy that the diametral creep of the Zr-2.5Nb pressure tube is governed by the texture. The aim of this work is to elucidate the governing factor of creep of the Zr-2.5Nb tubes and to find a way of making improved Zr-2.5Nb pressure tubes with a lower diametral creep and axial elongation. To this end, we scrutinized Holt's experiment where the in-reactor creep behaviors of the Zr-2.5Nb micro-pressure tube (MPT) with a circumferential texture was compared with that of the Zr-2.5Nb fuel sheath (FS) with a radial texture. Accounting for the fact that thermal creep of Zr-2.5Nb alloy is affected by the Nb concentration in the {beta}-Zr, we demonstrate that the reduced creep is not dictated by the circumferential texture but by the increased Nb concentration in the {alpha}-Zr. This study suggests that the optimized manufacturing procedure of the Zr-2.5Nb tube would improve their in-reactor performances, extending their design life to over 30 years when compared to that of the current design of the

  20. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  1. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  2. A study on the application of CRUDTRAN code in primary systems of domestic pressurized heavy-water reactors for prediction of radiation source term

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jong Soon; Cho, Hoon Jo; Jung, Min Young; Lee, Sang Heon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

  3. 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)

  4. 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)

  5. 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.

  6. Co-gasification of different rank coals with biomass and petroleum coke in a high-pressure reactor for H(2)-rich gas production.

    Science.gov (United States)

    Fermoso, J; Arias, B; Gil, M V; Plaza, M G; Pevida, C; Pis, J J; Rubiera, F

    2010-05-01

    Four coals of different rank were gasified, using a steam/oxygen mixture as gasifying agent, at atmospheric and elevated pressure in a fixed bed reactor fitted with a solids feeding system in continuous mode. Independently of coal rank, an increase in gasification pressure led to a decrease in H(2) + CO production and carbon conversion. Gasification of the different rank coals revealed that the higher the carbon content and reactivity, the greater the hydrogen production. Co-gasification experiments of binary (coal-biomass) and ternary blends (coal-petcoke-biomass) were conducted at high pressure to study possible synergetic effects. Interactions between the blend components were found to modify the gas production. An improvement in hydrogen production and cold gas efficiency was achieved when the coal was gasified with biomass. Copyright 2009 Elsevier Ltd. All rights reserved.

  7. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    Science.gov (United States)

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Lee-Taylor, J.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-11-01

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3, and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistry in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from volatile organic compounds (VOCs), NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach 2 orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s-1), 10-fold OH suppression is observed at OHRext ~ 100 s-1, which is similar or lower than used in many laboratory studies. The range of modeled OH suppression for literature experiments is consistent with the measured values except for those with isoprene. The finding on OH suppression may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over a 1-order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross-sections in the

  8. Estimation of fracture toughness of Zr 2.5% Nb pressure tube of Pressurised Heavy Water Reactor using cyclic ball indentation technique

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.; Rama Rao, A.

    2016-08-15

    Highlights: • Measurement of fracture toughness of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situ Property Measurement System (IProMS) has been designed in house. • Conventional and IProMS tests conducted on pressure tube spool pieces having different mechanical properties. • Correlation has been established between the conventional and IProMS estimated fracture properties. - Abstract: In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes made up of Zr 2.5 wt% Nb alloy. Pressure tubes undergo degradation during its service life due to high pressure, high temperature and radiation environment. Measurement of mechanical properties of degraded pressure tubes is important for assessing their fitness for further operation. Presently as per safety guidelines imposed by the regulatory body, a few pre-decided pressure tubes are removed from the reactor core at regular intervals during the planned reactor shut down to carry out post irradiation examination (PIE) in a laboratory which consumes lots of man-rem and imposes economic penalties. Hence a system is indeed felt necessary which can carry out experimental trials for measurement of mechanical properties of pressure tubes under in situ conditions. The only way to accomplish this important objective is to develop a system based on an in situ measurement technique. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing an indentation test either on the outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ conditions. Considering the importance of such measurements, an In situ Property

  9. A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions

    Energy Technology Data Exchange (ETDEWEB)

    Morfin, Franck; Piccolo, Laurent [Institut de recherches sur la catalyse et l' environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

    2013-09-15

    A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup −4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

  10. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  11. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

    1997-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  12. Pressurized pyrolysis of rice husk in an inert gas sweeping fixed-bed reactor with a focus on bio-oil deoxygenation.

    Science.gov (United States)

    Qian, Yangyang; Zhang, Jie; Wang, Jie

    2014-12-01

    The pyrolysis of rice husk was conducted in a fixed-bed reactor with a sweeping nitrogen gas to investigate the effects of pressure on the pyrolytic behaviors. The release rates of main gases during the pyrolysis, the distributions of four products (char, bio-oil, water and gas), the elemental compositions of char, bio-oil and gas, and the typical compounds in bio-oil were determined. It was found that the elevation of pressure from 0.1MPa to 5.0MPa facilitated the dehydration and decarboxylation of bio-oil, and the bio-oils obtained under the elevated pressures had significantly less oxygen and higher calorific value than those obtained under atmospheric pressure. The former bio-oils embraced more acetic acid, phenols and guaiacols. The elevation of pressure increased the formation of CH4 partially via the gas-phase reactions. An attempt is made in this study to clarify "the pure pressure effect" and "the combined effect with residence time". Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Flow Distribution and Pressure Loss in Subchannels of a Wire-Wrapped 37-pin Rod Bundle for a Sodium-Cooled Fast Reactor

    OpenAIRE

    Chang, Seok-Kyu; Euh, Dong-Jin; Choi, Hae Seob; Kim, Hyungmo; Choi, Sun Rock; Lee, Hyeong-Yeon

    2016-01-01

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20–115% ...

  14. Statistical analysis of the blowdown phase of a loss-of-coolant accident in a pressurized water reactor as calculated by RELAP4/MOD6

    Energy Technology Data Exchange (ETDEWEB)

    Berman, M.; Byers, R.K.; Steck, G.P.

    1979-01-01

    A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators.

  15. Gd/sub 2/O/sub 3/ up to 9 weight percent, an established burnable poison for advanced fuel management in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boehm, W.; Kiehlmann, H.D.; Neufert, A.; Peehs, M.

    1987-07-01

    High weight percent Gd/sub 2/O/sub 3/ has given excellent results when applied as burnable poison in pressurized water reactors for advanced fuel management tasks. Poisoning of up to 9 weight% Gd/sub 2/O/sub 3/ has been implemented in commercial reload cores to match the requirements of full low leakage loading and cycle extension strategies. Operational performance has confirmed that the high degree of accuracy achieved for calculational methods for standard loading applications also applies for highly Gd poisoned cores. The UO/sub 2/-Gd/sub 2/O/sub 3/ fabrication process has been rationalized by the use of direct pelletizing.

  16. Propellant actuated nuclear reactor steam depressurization valve

    Science.gov (United States)

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  17. Two-phase flow experiments on Counter-Current Flow Limitation in a model of the hot leg of a pressurized water reactor (2015 test series)

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Matthias; Lucas, Dirk; Pietruske, Heiko; Szalinski, Lutz

    2016-12-15

    Counter-Current Flow Limitation (CCFL) is of importance for PWR safety analyses in several accident scenarios connected with loss of coolant. Basing on the experiences obtained during a first series of hot leg tests now new experiments on counter-current flow limitation were conducted in the TOPFLOW pressure vessel. The test series comprises air-water tests at 1 and 2 bar as well as steam-water tests at 10, 25 and 50 bar. During the experiments the flow structure was observed along the hot leg model using a high-speed camera and web-cams. In addition pressure was measured at several positions along the horizontal part and the water levels in the reactor-simulator and steam-generator-simulator tanks were determined. This report documents the experimental setup including the description of operational and special measuring techniques, the experimental procedure and the data obtained. From these data flooding curves were obtained basing on the Wallis parameter. The results show a slight shift of the curves in dependency of the pressure. In addition a slight decrease of the slope was found with increasing pressure. Additional investigations concern the effects of hysteresis and the frequencies of liquid slugs. The latter ones show a dependency on pressure and the mass flow rate of the injected water. The data are available for CFD-model development and validation.

  18. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  19. HORIZONTAL BOILING REACTOR SYSTEM

    Science.gov (United States)

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  20. 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.

  1. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  2. Simulation of the spatial distribution of the acoustic pressure in sonochemical reactors with numerical methods: a review.

    Science.gov (United States)

    Tudela, Ignacio; Sáez, Verónica; Esclapez, María Deseada; Díez-García, María Isabel; Bonete, Pedro; González-García, José

    2014-05-01

    Numerical methods for the calculation of the acoustic field inside sonoreactors have rapidly emerged in the last 15 years. This paper summarizes some of the most important works on this topic presented in the past, along with the diverse numerical works that have been published since then, reviewing the state of the art from a qualitative point of view. In this sense, we illustrate and discuss some of the models recently developed by the scientific community to deal with some of the complex events that take place in a sonochemical reactor such as the vibration of the reactor walls and the nonlinear phenomena inherent to the presence of ultrasonic cavitation. In addition, we point out some of the upcoming challenges that must be addressed in order to develop a reliable tool for the proper designing of efficient sonoreactors and the scale-up of sonochemical processes. Copyright © 2013 Elsevier B.V. All rights reserved.

  3. The Hydrodynamic Characteristics of Cocurrent Downflow and Cocurrent Upflow Gas-Liquid-Solid Catalytic Fixed Bed Reactors: the Effect of Pressure Les caractéristiques hydrodynamiques des réacteurs gaz-liquide-solide à lit de catalyseur fixe à écoulement cocourant montant et descendant : l'influence de la pression

    OpenAIRE

    Wild G.; Larachi F.; Laurent A.

    2006-01-01

    While most catalytic fixed bed gas-liquid reactors of the petrol industry work at quite high pressures, the academic scientific work in this field concerned itself almost exclusively with the domain of approximatively atmospheric pressures. The authors present the results of some years of experimental investigations on the hydrodynamic characteristics of trickle bed reactors and lately of cocurrent upflow reactors. During the last years, results were also obtained under pressures up to 8 MPa....

  4. Tailoring nanomaterial products through electrode material and oxygen partial pressure in a mini-arc plasma reactor

    Science.gov (United States)

    Cui, Shumao; Mattson, Eric C.; Lu, Ganhua; Hirschmugl, Carol; Gajdardziska-Josifovska, Marija; Chen, Junhong

    2012-03-01

    Nanomaterials with controllable morphology and composition are synthesized by a simple one-step vapor condensation process using a mini-arc plasma source. Through systematic investigation of mini-arc reactor parameters, the roles of carrier gas, electrode material, and precursor on producing diverse nanomaterial products are revealed. Desired nanomaterial products, including tungsten oxide nanoparticles (NPs), tungsten oxide nanorods (NRs), tungsten oxide and tin oxide NP mixtures and pure tin dioxide NPs can thus be obtained by tailoring reaction conditions. The amount of oxygen in the reactor is critical to determining the final nanomaterial product. Without any precursor material present, a lower level of oxygen in the reactor favors the production of W18O49 NRs with tungsten as cathode, while a high level of oxygen produces more round WO3 NPs. With the presence of a precursor material, amorphous particles are favored with a high ratio of argon:oxygen. Oxygen is also found to affect tin oxide crystallization from its amorphous phase in the thermal annealing. Results from this study can be used for guiding gas phase nanomaterial synthesis in the future.

  5. High Temperature Humidity Sensor for Detection of Leak Through Slits and Cracks in Pressurized Nuclear Power Reactor Pipes

    Directory of Open Access Journals (Sweden)

    Debdulal Saha

    2007-03-01

    Full Text Available The leak before break (LBB concept is well known to nuclear power reactor. The problem is common to water power reactor. This is based on the premise that a detectable leak will develop before catastrophic break occurs. The main purpose of the present study is to develop tape cast MgCr2O4+35mole% TiO2 and gel cast g-Al2O3 humidity sensor for use in LBB applications at 3000C. The material capacitance varies with transient injection of water vapour adsorption. In actual plant, the sensors are placed on a steam pipe surrounded by heat insulation. The pipe unites the nuclear reactor and power generator. The analysis of humidity distribution in the annulus is calculated assuring leak rate 0.1gpm in a 30 m long tube. In this paper, analysis is done on the basis of the two types of sensor using AC frequency. Performance characteristics are observed for the LLB application.

  6. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  7. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  8. 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)

  9. Chemical detoxification of trichloroethylene and 1,1,1-trichloroethane in a microwave discharge plasma reactor at atmospheric pressure

    Energy Technology Data Exchange (ETDEWEB)

    Krause, T.R.; Helt, J.E.

    1991-12-31

    This report focuses on the application of plasma technology to hazardous waste treatment. Microwave sustained plasmas are used to thermal degrade trichloroethylene and trichloroethane at atmospheric pressure. (JL)

  10. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  11. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India. PACS No. 28.41.−i. 1. ... water reactors, mainly pressurized heavy water reactors (PHWRs) to extract ∼10 GWe capacity for ..... commissioning phase and most of the supporting systems have been commissioned and.

  12. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  13. Inter-comparison of JEF-2.2 and JEFF-3.1 evaluated nuclear data through Monte Carlo analysis of VVER-1000 MOX Core Computational Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L., E-mail: thilagam@igcar.gov.i [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India); Karthikeyan, R., E-mail: rkarthi@barc.gov.i [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Jagannathan, V., E-mail: v_jagan1952@rediffmail.co [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Subbaiah, K.V.; Lee, S.M. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2010-02-15

    The nuclear data forms a vital component in reactor core physics computations. The nuclear data is evaluated and modified on a continuous basis by different nuclear data centres and laboratories worldwide. The work on upgradation of the nuclear data is being carried out using new evaluations obtained through experiments and theoretical models to enhance their accuracy. Use of different sets of cross-section data in the analysis of a benchmark problem is a source of strong feedback for further improvements in data by mutual comparison of results. These comparisons also help to find out the best evaluated cross-section data released. Towards this objective, an attempt has been made to inter-compare JEF-2.2 and JEFF-3.1 evaluated nuclear data through the Monte Carlo simulation of 'VVER-1000 MOX Core Computational Benchmark'. This study deals with the calculation and inter-comparison of reactor parameters such as multiplication factors, cell average and assembly average fission reaction rate distributions estimated for various reactor state descriptions specified in the benchmark. Point-wise cross-section libraries processed from the JEF-2.2 and JEFF-3.1 evaluated data are used in the analysis. Concerning the multiplication factors and fission rate distributions, considerable differences are observed between the two libraries. While performing the MCNP calculations with JEFF-3.1 data, it is observed that the deviations of effective neutron multiplication factors (k{sub eff}) from those of benchmark standard MCU results are lower by about approx0.100% for the most of the states than those computed using JEF-2.2. Fission rate distributions using JEFF-3.1 data are also found to have significant deviations up to +-9.2% compared to calculations with its earlier version JEF-2.2 data. Some interesting trends on the used nuclear data are identified from the discrepancies of the individual results. The cause for considerable changes in the calculated parameters are

  14. 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.

  15. FABRICATION OF CNTS BY TOLUENE DECOMPOSITION IN A NEW REACTOR BASED ON AN ATMOSPHERIC PRESSURE PLASMA JET COUPLED TO A CVD SYSTEM

    Directory of Open Access Journals (Sweden)

    FELIPE RAMÍREZ-HERNÁNDEZ

    2017-03-01

    Full Text Available Here, we present a method to produce carbon nanotubes (CNTs based on the coupling between two conventional techniques used for the preparation of nanostructures: an arc-jet as a source of plasma and a chemical vapour deposition (CVD system. We call this system as an “atmospheric pressure plasma (APP-enhanced CVD” (APPE-CVD. This reactor was used to grow CNTs on non-flat aluminosilicate substrates by the decomposition of toluene (carbon source in the presence of ferrocene (as a catalyst. Both, CNTs and by-products of carbon were collected at three different temperatures (780, 820 and 860 °C in different regions of the APPE-CVD system. These samples were analysed by thermogravimetric analysis (TGA and DTG, scanning electron microscopy (SEM and Raman spectroscopy in order to determine the effect of APP on the thermal stability of the as-grown CNTs. It was found that the amount of metal catalyst in the synthesised CNTs is reduced by applying APP, being 820 °C the optimal temperature to produce CNTs with a high yield and carbon purity (95 wt. %. In contrast, when the synthesis temperature was fixed at 780 °C or 860 °C, amorphous carbon or CNTs with different structural defects, respectively, was formed through APEE-CVD reactor. We recommended the use of non-flat aluminosilicate particles as supports to increase CNT yield and facilitate the removal of deposits from the substrate surface. The approach that we implemented (to synthesise CNTs by using the APPE-CVD reactor may be useful to produce these nanostructures on a gram-scale for use in basic studies. The approach may also be scaled up for mass production.

  16. Networked Control of a Large Pressurized Heavy Water Reactor (PHWR) With Discrete Proportional-Integral-Derivative (PID) Controllers

    Science.gov (United States)

    Dasgupta, Soumya; Routh, Avijit; Banerjee, Shohan; Agilageswari, K.; Balasubramanian, R.; Bhandarkar, S. G.; Chattopadhyay, Sujit; Kumar, Manoj; Gupta, Amitava

    2013-10-01

    Networked control has been attempted for a wide variety of plants using memory-less State Feedback controllers. Networked Control Systems (NCSs) have also been mooted for PHWRs and implemented in practical systems e.g. a Reactor Regulating System (RRS) of the Indian PHWRs. While existing approaches using State Feedback controllers guarantee stability under data loss and variable delay which are inherent in real time communications, they cannot be extended to discrete PID controllers which involve a memory element. In this paper, a switched system formulation has been used to analyze the stability of a nuclear reactor with networked discrete PID controllers in the RRS under specified data loss and variable delay. The results have been verified using MATLAB xPC target based Real-time Simulation involving multiple simulation nodes integrated over switched Ethernet. A Distributed Network Analyzer (DNA) has been used to log actual network traffic to establish the credibility of the assumptions related to data loss and variable delay in communication.

  17. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany); Gillemot, F. [Centre for Energy Research of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Hernández-Mayoral, M.; Serrano, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Török, G. [Wigner Research Center for Physics of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Ulbricht, A.; Altstadt, E. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany)

    2015-06-15

    Highlights: • TEM and SANS were applied to estimate mean size and number density of loops, nanovoids and Cu-rich clusters. • A three-feature dispersed-barrier hardening model was applied to estimate the yield stress increase. • The values and errors of the dimensionless obstacle strength were estimated in a consistent way. • Nanovoids are stronger obstacles for dislocation glide than dislocation loops, loops are stronger than Cu-rich clusters. • For reactor-relevant conditions, Cu-rich clusters contribute most to hardening due to their high number density. - Abstract: 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.

  18. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  19. 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.

  20. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  1. Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data

    Energy Technology Data Exchange (ETDEWEB)

    McAfee, W.J.; Bass, B.R.; Bryson, J.W.

    1998-07-01

    A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

  2. 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.

  3. 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.

  4. J-R Fracture Resistance of SA533 Gr.B-Cl.1 Steel for Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji-Hyun; Hong, Seokmin; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A rolled plate might show different mechanical behaviors from a forging, even though they contain same chemical compositions. Furthermore, it is known that the fracture behavior of a rolled plate is very sensitive to material orientation comparing to a forging. In this study, the J-R fracture resistances of SA533 Gr.B-Cl.1 plate were measured at reactor operating temperature and the material orientation sensitivity was discussed. The decrease of fracture resistance of this kind of low alloy steel at an elevated temperature is known as the effect of dynamic strain aging (DSA). It was attributed to that the carbides and grains elongated to primary rolling direction, so that the aspect ratio of carbides and grains in the specimen with T-L orientation is larger. Generally, the hard second phase could take a roll of trigger point of unstable fracture. It is needed that the fracture surfaces of the tested specimens to be examined profoundly.

  5. STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

    Directory of Open Access Journals (Sweden)

    A. V. Korolyev

    2017-01-01

    Full Text Available The research presented in the article conforms to the severe accident that took place at the Three Mail Island nuclear power plant in the USA. The research is focused on improving the reliability of the pressure compensator that is an important equipment of the primary circuit. In order to simulate such a situation, the stand has been developed to simulate the design of the pressurizer of the PWR-440 reactor, in particular an elliptical shape of the upper lid which has a steam outlet pipe at the top of the construction that creates conditions for occurrence of such water hammers. For the experiments, an installation has been created that makes it possible to measure and record the water hammering that occur when the tanks are filled. Measurement of the amplitude of the water hammering was carried out by a specially developed piezoelectric sensor, and the registration – by a light-beam oscilloscope. The technique of carrying out the experiment is described and the results of an experimental study of the water hammers arising when the vessels are completely filled are presented. Quantitative data were obtained on the amplitudes of the hydraulic impacts depending on the rate of filling of the vessel and the diameter of the outlet, the maximum pressure of the hydraulic shock was 7–9 atm. Comparison of calculated and experimental data has been performed. The allowable discrepancy is explained by the calculated value of the system stiffness coefficient, which did not take into account the presence of welded seams in the tank that imparts the system with additional rigidity. The calculated relationships are obtained, that make it possible to estimate the amplitudes of the water hammers through the acceleration of the water level in front of the outlet from a vessel with an elliptical bottom. The possibility of a water hammer in the pressure compensator is demonstrated by experiment and by theoretical calculations. Based on the experimental data, a

  6. The impact of ODA microadditions into secondary system on corrosion rate reduction in VVER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Avdeev, A.A.; Kukushkin, A.N.; Repin, D.A. [All-Russia Research and Design Inst. of Nuclear Power Machine Building (VNIIAM), Moscow (Russian Federation); Omelchuk, V.V.; Barmin, L.F. [Kola Nuclear Power Plant, Polyarnye Zori, Murmansk region (Russian Federation); Yurmanov, V.A. [N.A. Dollezhal Research and Development Inst. of Power Engineering (NIKIET), Moscow (Russian Federation); Czempik, E. [RECON GmbH, Leipzig (Germany)

    2010-07-01

    Injection of film-forming corrosion inhibitors is a challenging way of suppressing erosion-corrosion and crud induced corrosion in power stations. Films of surface-active inhibitors, such as octadecylamine (ODA) provide a diffusion barrier to penetration of corrosion-aggressive ions onto the metal surface. Erosion and corrosion tests were conducted in autoclaves and on a pilot steam generator (SG) design to look into the impact of ODA. To accelerate corrosion process tests were conducted in a more aggressive environment as compared to actual operating conditions, including high chloride concentration and stress levels. It is not only important to reduce deposition growth, but also to wash out deposits previously formed on heat exchanger surfaces. This allows to reduce the risk of local corrosion and corrosion cracking development. A number of VVER plants have conducted full-scale testing that confirmed the impact of ODA microadditions on local corrosion mitigation. Some PWR plants are testing injection of surface-active dispersants to loosen SG deposits. Multiple studies proved ODA ability to remove chlorides from smooth surfaces which allowed to reduce the rate of microcrack growth. Trial testing has shown that the rate of corrosion cracking on SG tubes was reduced by 60-70% owing to ODA injections. Such effect was due to significant reduction in chlorides absorption by the metal surface during the year of ODA injection. Tests on a pilot SG design have shown that ODA could be used for partially wash out deposits from a heating surface. This also minimizes local corrosion. The tests showed that ODA microadditions remove chlorides from microcracks and crevices on SG tubing spacer grids. The ability to wash out previously formed deposits allows to reduce risk of local corrosion and cracking. The abilities of microadditions of film-forming corrosion inhibitors identified through the above mentioned testing could be used under the VVER plant life extension program. The

  7. Pressure Drop and Catalytic Dehydrogenation of NaBH{sub 4} Solution Across Pin Fin Structures in a Microchannel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Moon [Korea Institute of Industrial Technology, Cheonan (Korea, Republic of); Choi, Seok Hyun [Key Valve Technologies Ltd., Siheung (Korea, Republic of); Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of)

    2017-06-15

    Dehydrogenation from the hydrolysis of a sodium borohydride (NaBH{sub 4}) solution has been of interest owing to its high theoretical hydrogen storage capacity (10.8 wt.%) and potentially safe operation. An experimental study has been performed on the catalytic reaction rate and pressure drop of a NaBH4 solution over both a single microchannel with a hydraulic diameter of 300 μm and a staggered array of micro pin fins in the microchannel with hydraulic diameter of 50 μm. The catalytic reaction rates and pressure drops were obtained under Reynolds numbers from 1 to 60 and solution concentrations from 5 to 20 wt.%. Moreover, reacting flows were visualized using a high-speed camera with a macro zoom lens. As a result, both the amount of hydrogenation and pressure drop are 2.45 times and 1.5 times larger in a pin fin microchannel array than in a single microchannel, respectively.

  8. 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.

  9. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  10. Ductile-Brittle Transition Behavior in Tempered Martensitic 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)

    2010-05-15

    Reactor pressure vessels (RPVs) operate under severe conditions of elevated temperature, high pressure, and irradiation. Therefore, a combination of sufficient strength, toughness, good weldability, and high irradiation resistance are required for RPV materials. SA508 Gr.4N low alloy steel, which has higher Ni and Cr contents than those of commercial RPV steel, Gr.3 steel, is considered as a candidate material due to its excellent mechanical properties from tempered martensitic microstructure. The ferritic steels such as Gr.3 and Gr.4N low alloy steels reveal a ductile-brittle transition and large scatters in the fracture toughness within a small temperature range. Recently, there are some observations of the steeper transition behavior in the tempered martensitic steels, such as Eurofer97 than the transition behavior of commercial RPV steels. It was also reported that the fracture toughness increased discontinuously when the phase fraction of the tempered martensite was over a critical fraction in the heat affected zones of SA508 Gr.3. Therefore, it may be necessary to evaluate the changes of transition behavior with a microstructure for the tempered martensitic SA508 Gr.4N low alloy steel. In this study, the fracture toughness for SA508 Gr.4N low alloy steels was evaluated from a view point of the temperature dependency with phase fraction of tempered martensite controlled by cooling rate. Additionally, a possible modification of the fracture toughness master curve was proposed and discussed

  11. Cooperation in simulating a pressurized coal dust reactor with different simulation programs; Kooperation bei der Simulation eines Druckkohlenstaubreaktors mit unterschiedlichen Simulationsprogrammen

    Energy Technology Data Exchange (ETDEWEB)

    Ayar, A.; Hecken, M. [Technische Hochschule Aachen (Germany). Lehrstuhl fuer Waermeuebertragung und Klimatechnik; Mohr, M.; Murza, S. [Bochum Univ. (Germany). Lehrstuhl fuer Energieanlagentechnik; Richter, S.; Stroehle, J. [Stuttgart Univ. (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen

    1999-09-01

    The contribution presents the results of a cooperative project of three different universities - Department of Heat Transfer and Air Conditioning Engineering (WUeK) of RWTH Aachen, Department of Power Sytems Engineering (LEAT) of Ruhr-Universitaet Bochum, Institute of Chemical Engineering and Steam Boiler Engineering (IVD) of Stuttgart University - which involved modellilng a benchmark flame of the pressurized coal dust reactor of the Departmentof Heat Transfer and Air Conditioning Engineering. The conditions of the benchmark flame reflect a real firnace condition which so far has never been investigated experimentally. The contribution compares the results of the three different simulation programs and presents an outlook to the necessary further developments of the programs for modelling plants for pressurized coal dust combustion. [Deutsch] Es werden die Ergebnisse der Zusammenarbeit der drei beteiligten Institute - Lehrstuhl fuer Waermeuebertragung und Klimatechnik (WUeK) der RWTH Aachen, Lehrstuhl fuer Energieanlagentechnik (LEAT) der Ruhr-Uni Bochum und Institut fuer Verfahrenstechnik und Dampfkesselwesen (IVD) der Uni Stuttgart - bei der Modellierung einer Benchmarkflamme des Druckkohlenstaubreaktors am Lehrstuhl fuer Waermeuebertragung und Klimatechnik dargestellt. Diese Bedingungen fuer diese Benchmarkflamme sind einem realen, bis dato noch nicht experimentell untersuchten Feuerungszustand nachempfunden. Der Beitrag beinhaltet den Vergleich der mit drei unterschiedlichen Simulationsprogrammen erzielten Ergebnisse und gibt einen Ausblick auf die notwendigen Weiterentwicklungen der Programme hinsichtlich der Modellierung druckkohlenstaubgefeuerter Anlagen. (orig.)

  12. Flow Distribution and Pressure Loss in Subchannels of a Wire-Wrapped 37-pin Rod Bundle for a Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Seok-Kyu Chang

    2016-04-01

    Full Text Available A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20–115% of the nominal flow rate and 60°C (equivalent to Re ∼ 37,100 at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  13. Low cycle thermomechanical fatigue of reactor steels: Microstructural and fractographic investigations

    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); Kasl, Josef; Jandova, Dagmar [Výzkumný a zkušební ústav Plzeň s.r.o., Tylova 1581/46, 316 00 Plzen (Czech Republic); Jóni, Bertalan [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Eötvös Loránd University, Egyetem tér 1-3, Budapest H-1053 (Hungary); Misják, Fanni [Centre for Energy Research, Institute of Technical Physics and Materials Science, Konkoly-Thege M. 29-33, Budapest H-1121 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-07-29

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of a VVER-440 reactor pressure vessel were investigated under fully reversed total strain controlled low cycle fatigue tests. The measurements were carried out in isothermal conditions at 260 °C and with thermal-mechanical conditions in the range 150–270 °C using a GLEEBLE-3800 servo-hydraulic thermal-mechanical simulator. The low cycle fatigue results were evaluated with the Coffin–Manson law, and the parameters of the Ramberg–Osgood stress–strain relation were investigated. Fracture mechanics behavior was observed using scanning electron microscopic analysis of the crack shapes and fracture surfaces. Crack propagation was assessed in relation to the actual crack size and the loading level. Interrupted fatigue tests were also carried out to investigate the kinetics of the fatigue evolution of the materials. Microstructural evaluation of the samples was performed using light, scanning and transmission electron microscopy as well as X-ray diffraction, and measurement of dislocations was completed using TEM and XRD. The course of dislocation density in relation to cumulative usage factor was similar for both steels. However, the nature and distribution of dislocations were different in the individual steels and this resulted in different mechanical behaviors. The nature of the fracture surfaces of both steels appeared similar despite differences in dislocation arrangement. The distances between striation lines initially increased with increasing crack length and then became saturated. The low cycle fatigue behavior investigated can provide a reference for the remaining life assessment and lifetime extension analysis of nuclear power plant components.

  14. Dynamics of the atmospheric pressure diffuse dielectric barrier discharge between cylindrical electrodes in roll-to-roll PECVD reactor

    Science.gov (United States)

    Starostin, Sergey A.; Welzel, Stefan; Liu, Yaoge; van der Velden-Schuermans, Bernadette; Bouwstra, Jan B.; van de Sanden, Mauritius C. M.; de Vries, Hindrik W.

    2015-07-01

    The high current diffuse dielectric barrier discharge (DBD) was operated in a bi-axial cylindrical electrode configuration using nitrogen, oxygen and argon gas flow with the addition of tetraethyl orthosilicate as precursor for silica-like film deposition. The behaviour of the transient plasma was visualized by means of fast imaging from two orthogonal directions. The formation and propagation (~3 × 104 m s-1) of lateral ionization waves with the transverse light emission structure similar to the low pressure glow discharge was observed at time scales below 1 µs. Despite plasma non-uniformity at nanosecond time scale the deposition process on the web-rolled polymer results in smooth well adherent films with good film uniformity and excellent gas diffusion barrier properties. Contribution to the topical issue "The 14th International Symposium on High Pressure Low Temperature Plasma Chemistry (HAKONE XIV)", edited by Nicolas Gherardi, Ronny Brandenburg and Lars Stollenwark

  15. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  16. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  17. 78 FR 34677 - Advisory Committee On Reactor Safeguards (ACRS); Meeting of the Acrs Subcommittee on Materials...

    Science.gov (United States)

    2013-06-10

    ... COMMISSION Advisory Committee On Reactor Safeguards (ACRS); Meeting of the Acrs Subcommittee on Materials, Metallurgy & Reactor Fuels; Notice of Meeting The ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels... pellet-cladding interaction during anticipated operational occurrences for Pressurized Water Reactors...

  18. The Westinghouse Small Modular Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Garcia, F. J.; Memmott, M. J.

    2012-07-01

    Westinghouse has developed a small modular reactor (SMR) which incorporates an integral pressurized water reactor (PWR) design. The Westinghouse SMR design also utilizes many of the key features and innovative concepts from the AP1000 plant, including passive safety systems.

  19. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Completion of the dismantling, cutting and packaging for final disposal of the MZFR reactor pressure vessel including internals; Abschluss der vollstaendigen Demontage, Zerlegung und endlagergerechten Verpackung des 400 t schweren MZFR-Reaktordruckbehaelters incl. Einbauten

    Energy Technology Data Exchange (ETDEWEB)

    Prechtl, E.; Eisenmann, B.; Weis, A. [Forschungszentrum Karlsruhe GmbH (DE). Hauptabt. Projekte (HAP-MZFR); Suessdorf, W. [Studsvik GmbH und Co. KG, Pforzheim (Germany); Loeb, A.; Stanke, D.; Thoma, M. [NIS Ingenieurgesellschaft mbH, Alzenau (Germany)

    2008-07-01

    The test reactor 400 tons MZFR reactor pressure vessel has been dismantled, dissected and packaged for final disposal. The authors describe the initial situation of the reactor after 19 years of operation. The dismantling was complicated by the high radioactivity and the high contamination due to the constricted space in the building. In 1995/1996 remote cutting tools for hot components were not yet available. Established techniques were adapted and refined for the special purpose. The dissected vessel was packaged for final disposal in the final repository Konrad. The authors describe the dismantling concept including boundary conditions and general requirements, the i