Atomic reactor thermal engineering

International Nuclear Information System (INIS)

Kim, Gwang Ryong

1983-02-01

This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

International Nuclear Information System (INIS)

Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

1994-08-01

We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

2015-12-01

Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

Energy Technology Data Exchange (ETDEWEB)

Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

2017-01-15

Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

The Choice of thermal reactor systems. A report by the National Nuclear Corporation Limited

International Nuclear Information System (INIS)

1977-01-01

The report to the Secretary of State in Great Britain by the National Nuclear Corporation following their assessment of the three thermal reactor systems, the AGR, PWR and SGHWR type reactors, which was performed in order to assist in the decision on the choice of thermal reactors for the U.K., is in three parts. Part I is an assessment of the three systems. It comprises: a description of the general method of assessment; a commentary in which are summarised discussions on the most important issues influencing reactor choice, i.e. safety, component failure, operational characteristics, development programme, construction programme; implications for the U.K. industry; costs; and reference design of each system. Part II consists of related questions and answers accompanied by commentaries on public acceptability and views from industry. Part III contains some conclusions including an analysis on the implications of the choices open and a summary of the main features of the assessment. (U.K.)

Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

International Nuclear Information System (INIS)

Kawahara, Toshio; Matsushita, Tadashi

1977-01-01

The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

International Nuclear Information System (INIS)

Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

1994-07-01

The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

Thermal-hydraulic analysis of nuclear reactors

CERN Document Server

Zohuri, Bahman

2015-01-01

This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

The dissolver paradox as a coupled fast-thermal reactor

International Nuclear Information System (INIS)

Lutz, H.F.; Webb, P.S.

1993-05-01

The dissolver paradox is treated as coupled fast-thermal reactors. Each reactor is sub-critical but the coupling is sufficient to form a critical system. The practical importance of the system occurs when the fast system by itself is mass limited and the thermal system by itself is volume limited. Numerous 1D calculations have been made to calculate the neutron multiplication parameters of the separate fast and thermal systems that occur in the dissolver paradox. A model has been developed to describe the coupling between the systems. Monte Carlo calculations using the MCNP code have tested the model

Calculation of the neutron parameters of fast thermal reactor

International Nuclear Information System (INIS)

Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

1975-01-01

The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

Test program for NIS calibration to reactor thermal output in HTTR

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

2000-03-01

Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

Development of demonstration advanced thermal reactor

Energy Technology Data Exchange (ETDEWEB)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

1982-08-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

Development of demonstration advanced thermal reactor

International Nuclear Information System (INIS)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

1982-01-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

Response of a thermal barrier system to acoustic excitation in a gas turbine nuclear reactor

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Blevins, R.D.

1980-11-01

A gas turbine located within a High-Temperature Gas-Cooled Reactor (HTGR) induces high acoustic sound pressure levels into the primary coolant (helium). This acoustic loading induces high cycle fatigue stresses which may control the design of the thermal barrier system. This study examines the dynamic response of a thermal barrier configuration consisting of a fibrous insulation compressed against the reactor vessel by a coverplate which is held in position by a central attachment fixture. The results of dynamic vibration analyses indicate the effect of the plate size and curvature and the attachment size on the response of the thermal barrier

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

International Nuclear Information System (INIS)

Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

2016-01-01

Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

2016-08-15

Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

Thorium utilisation in thermal reactors

International Nuclear Information System (INIS)

Balakrishnan, K.

1997-01-01

It is now more or less accepted that the best way to use thorium is in thermal reactors. This is due to the fact that U233 is a good material in the thermal spectrum. Studies of different thorium cycles in various reactor concepts had been carried out in the early days of nuclear power. After three decades of neglect, the world is once again looking at thorium with some interest. We in India have been studying thorium cycles in most of the existing thermal reactor concepts, with greater emphasis on heavy water reactors. In this paper, we report some of the work done in India on different thorium cycles in the Indian pressurized heavy water reactor (PHWR), and also give a description of the design of the advanced heavy water reactor (AHWR). (author). 1 ref., 2 tabs., 5 figs

Minor Actinide Burning in Thermal Reactors. A Report by the Working Party on Scientific Issues of Reactor Systems

International Nuclear Information System (INIS)

Hesketh, K.; Porsch, D.; Rimpault, G.; Taiwo, T.; Worrall, A.

2013-01-01

them to be considered candidates for transmutation. Fast reactors are needed to transmute TRUs because fast neutron cross sections are generally more effective in the fissioning of TRUs. However, studies have demonstrated that TRU transmutation rates can also be achieved in thermal reactors, although with serious limitations due to their accumulation through recycling and their impact on the safety of the plants. The transmutation of TRUs could potentially be carried out in many thermal reactors operating today, while waiting for a similar programme in fast reactors to allow commercial-scale operations in 20 to 30 years or more. Investment in fuel cycle plants could lead to even more efficient transmutation in fast reactors towards the end of the century. In the interim, the potential contribution of thermal reactors should not be overlooked. A considerable amount of research has already been conducted on minor actinide transmutation in thermal reactors, and the purpose of this report is to summarise the findings of this research. The report concentrates on general conclusions related to thermal reactors and foregoes a lengthy examination of the more technical details. While a commercial-scale implementation programme for a specific reactor type will need to address a multitude of very specific questions, the objective of this report is to provide the broad understanding necessary to inform high-level strategy and decision making. Chapter 1 of the report provides an introduction to minor actinide nuclear properties and discusses some of the arguments in favour of minor actinide recycling. The introduction is not specific to thermal reactors but could apply to any nuclear system in general. Chapter 2 discusses the potential role of thermal reactors in minor actinide recycling; Chapter 3 looks at the various technical issues and challenges presented by minor actinide recycling; Chapter 4 examines fuel cycle issues; Chapter 5 presents implications for thermal reactor

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

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

1995-01-01

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

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

Science.gov (United States)

Jefferies, K. S.; Tew, R. C.

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

International Nuclear Information System (INIS)

Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

2010-10-01

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

The role of heater thermal response in reactor thermal limits during oscillartory two-phase flows

Energy Technology Data Exchange (ETDEWEB)

Ruggles, A.E.; Brown, N.W. [Univ. of Tennessee, Knoxville, TN (United States); Vasil`ev, A.D. [Nuclear Safety Institute, Moscow, (Russian Federation); Wendel, M.W. [Oak Ridge National Lab., TN (United States)

1995-09-01

Analytical and numerical investigations of critical heat flux (CHF) and reactor thermal limits are conducted for oscillatory two-phase flows often associated with natural circulation conditions. It is shown that the CHF and associated thermal limits depend on the amplitude of the flow oscillations, the period of the flow oscillations, and the thermal properties and dimensions of the heater. The value of the thermal limit can be much lower in unsteady flow situations than would be expected using time average flow conditions. It is also shown that the properties of the heater strongly influence the thermal limit value in unsteady flow situations, which is very important to the design of experiments to evaluate thermal limits for reactor fuel systems.

Thermal reactor strategy

International Nuclear Information System (INIS)

1981-01-01

This statement sets down briefly the CEGB's views on the requirement for nuclear power and outlines current progress in the implementation of the CEGB's thermal reactor strategy. The programme is traced historically, together with statements of Government policy. The place of Magnox, AGR, SGHWR, PWR and fast breeder reactors is discussed. Advantages and problems associated with the various types are outlined. (U.K.)

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

International Nuclear Information System (INIS)

Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

1999-09-01

A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

Simulation study of multi-step model algorithmic control of the nuclear reactor thermal power tracking system

International Nuclear Information System (INIS)

Shi Xiaoping; Xu Tianshu

2001-01-01

The classical control method is usually hard to ensure the thermal power tracking accuracy, because the nuclear reactor system is a complex nonlinear system with uncertain parameters and disturbances. A sort of non-parameter model is constructed with the open-loop impulse response of the system. Furthermore, a sort of thermal power tracking digital control law is presented using the multi-step model algorithmic control principle. The control method presented had good tracking performance and robustness. It can work despite the existence of unmeasurable disturbances. The simulation experiment testifies the correctness and effectiveness of the method. The high accuracy matching between the thermal power and the referenced load is achieved

Methodology of Supervision by Analysis of Thermal Flux for Thermal Conduction of a Batch Chemical Reactor Equipped with a Monofluid Heating/Cooling System

Directory of Open Access Journals (Sweden)

Ghania Henini

2012-01-01

Full Text Available We present the thermal behavior of a batch reactor to jacket equipped with a monofluid heating/cooling system. Heating and cooling are provided respectively by an electrical resistance and two plate heat exchangers. The control of the temperature of the reaction is based on the supervision system. This strategy of management of the thermal devices is based on the usage of the thermal flux as manipulated variable. The modulation of the monofluid temperature by acting on the heating power or on the opening degrees of an air-to-open valve that delivers the monofluid to heat exchanger. The study shows that the application of this method for the conduct of the pilot reactor gives good results in simulation and that taking into account the dynamics of the various apparatuses greatly improves ride quality of conduct. In addition thermal control of an exothermic reaction (mononitration shows that the consideration of heat generated in the model representation improve the results by elimination any overshooting of the set-point temperature.

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

Nuclear reactor vessel fuel thermal insulating barrier

Science.gov (United States)

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

2013-03-19

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

Thermal and flow design of helium-cooled reactors

International Nuclear Information System (INIS)

Melese, G.; Katz, R.

1984-01-01

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

2015-12-31

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Reactor Thermal Hydraulic Numerical Calculation And Modeling

International Nuclear Information System (INIS)

Duong Ngoc Hai; Dang The Ba

2008-01-01

In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

Studies of thermal-hydraulics and plant systems for actinide burning fast reactor concept

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Misumi, Masahiro; Izaki, Makoto; Koike, Hiroyuki; Tanaka, Ryokichi

1984-01-01

As one of the methods to dispose long life actinide nuclides, the actinide burning fast reactor using only actinide wastes as the fuel has been proposed. Kawasaki Heavy Industries Ltd. carried out the conceptual examination on the ABFR cooled with helium gas, cooperating with Japan Atomic Energy Research Institute, and its feasibility and problems were clarified. In this report, the setting-up of various fundamental dimensions by the parameter survey of the thermal and flowing performance of the core, the examination of the thermal and flowing characteristics of the core based on the detailed power distribution, and the examination of the plant system centering around the main cooling system are outlined. The fuel is composed of actinide oxide and diluent MgO. The diluent is used for obtaining proper excess reactivity, and MgO has been taken up also in foreign countries, considering the compatibility with actinide oxide, the easiness of reprocessing and manufacture. The fuel element is of pin type, and actinide oxide and MgO pellets are in a SUS 316 cladding tube. This ABFR can treat the wastes from ten 1000 MWe power reactors, and has the power output of about 1000 MWt. (Kako, I.)

Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

International Nuclear Information System (INIS)

Tuomisto, Harri.

1987-06-01

In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.

2015-01-01

Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

Thermal Hydraulic Tests for Reactor Core Safety

Energy Technology Data Exchange (ETDEWEB)

Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

2007-06-15

The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

Thermal-hydraulics for space power, propulsion, and thermal management system design

International Nuclear Information System (INIS)

Krotiuk, W.J.

1990-01-01

The present volume discusses thermal-hydraulic aspects of current space projects, Space Station thermal management systems, the thermal design of the Space Station Free-Flying Platforms, the SP-100 Space Reactor Power System, advanced multi-MW space nuclear power concepts, chemical and electric propulsion systems, and such aspects of the Space Station two-phase thermal management system as its mechanical pumped loop and its capillary pumped loop's supporting technology. Also discussed are the startup thaw concept for the SP-100 Space Reactor Power System, calculational methods and experimental data for microgravity conditions, an isothermal gas-liquid flow at reduced gravity, low-gravity flow boiling, computations of Space Shuttle high pressure cryogenic turbopump ball bearing two-phase coolant flow, and reduced-gravity condensation

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

2004-02-01

To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

Thermal performances of an insulating structure for a reactor vessel

International Nuclear Information System (INIS)

Aranovitch, E.; Crutzen, S.; Le Det, M.; Denis, R.

1974-12-01

This report describes the thermal and technological tests performed on a multilayer thermal insulation system for high temperature gas reactors. It includes the description of test facilities, global tests, interpretation of data, and technological tests. Results obtained make it possible to predetermine with a satisfactory precision thermal performances under various nominal conditions

Nuclear reactor safety system

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1983-01-01

The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

The unique safety challenges of space reactor systems

International Nuclear Information System (INIS)

Lanes, S.J.; Marshall, A.C.

1991-01-01

Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

Energy Technology Data Exchange (ETDEWEB)

Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

2015-08-15

Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Komen, E M.J.; Bogaard, J.P.A. van den

1995-06-01

In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

Experiment study on thermal mixing performance of HTR-PM reactor outlet

Energy Technology Data Exchange (ETDEWEB)

Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-09-15

A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

International Nuclear Information System (INIS)

Finch, D.R.

1965-01-01

1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

SRAC: JAERI thermal reactor standard code system for reactor design and analysis

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

1983-01-01

The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

International Nuclear Information System (INIS)

Ninokata, H.; Kamide, H.

2011-01-01

In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Honda, T.; Maki, K.; Okazaki, T.

1994-01-01

Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors

International Nuclear Information System (INIS)

Gupta, S.K.; Kumar, Rajesh; Gaikwad, A.J.; Majumdar, P.; Agrawal, R.A.

2004-01-01

Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors (220 MWe) the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper. (author)

Analysis of thermal fatigue events in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

Adaptability of Brayton cycle conversion systems to fast, epithermal and thermal spectrum space nuclear reactors

International Nuclear Information System (INIS)

Tilliette, Z.P.

1988-01-01

The two French Government Agencies C.N.E.S. (Centre National d'Etudes Spatiales) and C.E.A. (Commissariat a l'Energie Atomique) are carrying out joint preliminary studies on space nuclear power systems for future ARIANE 5 launch vehicle applications. The Brayton cycle is the reference conversion system, whether the heat source is a liquid metal-cooled (NaK, Na or Li) reactor or a gas-cooled direct cycle concept. The search for an adequate utilization of this energy conversion means has prompted additional evaluations featuring the definition of satisfactory cycle conditions for these various kinds of reactor concepts. In addition to firstly studied fast and epithermal spectrum ones, thermal spectrum reactors can offer an opportunity of bringing out some distinctive features of the Brayton cycle, in particular for the temperature conditioning of the efficient metal hydrides (ZrH, Li/sub 7/H) moderators. One of the purposes of the paper is to confirm the potential of long lifetime ZrH moderated reactors associated with a gas cycle and to assess the thermodynamical consequences for both Nak(Na)-cooled or gas-cooled nuclear heat sources. This investigation is complemented by the definition of appropriate reactor arrangements which could be presented on a further occasion

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

Thermal-hydraulics of actinide burner reactors

International Nuclear Information System (INIS)

Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.

1989-07-01

As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2015-05-15

Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

Investigation of effective factors of transient thermal stress of the MONJU-System components

Energy Technology Data Exchange (ETDEWEB)

Inoue, Masaaki; Hirayama, Hiroshi; Kimura, Kimitaka; Jinbo, M. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

1999-03-01

Transient thermal stress of each system Component in the fast breeder reactor is an uncertain factor on it's structural design. The temperature distribution in a system component changes over a wide range in time and in space. An unified evaluation technique of thermal, hydraulic, and structural analysis, in which includes thermal striping, temperature stratification, transient thermal stress and the integrity of the system components, is required for the optimum design of tho fast reactor plant. Thermal boundary conditions should be set up by both the transient thermal stress analysis and the structural integrity evaluation of each system component. The reasonable thermal boundary conditions for the design of the MONJU and a demonstration fast reactor, are investigated. The temperature distribution analysis models and the thermal boundary conditions on the Y-piece structural parts of each system component, such as reactor vessel, intermediate heat exchanger, primary main circulation pump, steam generator, superheater and upper structure of reactor core, are illustrated in the report. (M. Suetake)

Analysis of Reactor Pressurized Thermal Shock Conditions Considering Upgrading of Systems Important to Safety

International Nuclear Information System (INIS)

Mazurok, A.S; Vyshemirskyij, M.P.

2015-01-01

The paper analyzes conditions of pressurized thermal shock on the reactor pressure vessel taking into account upgrading of the emergency core cooling system and primary overpressure protection system. For representative accident scenarios, calculation and comparative analysis was carried out. These scenarios include a small leak from the hot leg and PRZ SV stuck opening with re closure after 3600 sec and 3 SG heat transfer tube rupture. The efficiency of mass flow control by valves on the pump head (emergency core cooling systems) and cold overpressure protection (primary overpressure protection system) was analyzed. The thermal hydraulic model for RELAP5/Mod3.2 code with detailed downcomer (DC) model and changes in accordance with upgrades was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. The upgrades prevent excessive primary cooling and, consequently, help to preserve the RPV integrity and to avoid the formation of a through crack, which can lead to a severe accident

Flexible Conversion Ratio Fast Reactor Systems Evaluation

International Nuclear Information System (INIS)

Neil Todreas; Pavel Hejzlar

2008-01-01

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

Thermal characteristics analysis of microwaves reactor for pyrolysis of used cooking oil

Science.gov (United States)

Anis, Samsudin; Shahadati, Laily; Sumbodo, Wirawan; Wahyudi

2017-03-01

The research is objected to develop microwave reactor for pyrolysis of used cooking oil. The effect of microwave power as well as addition of char as absorber towards its thermal characteristic were investigated. Domestic microwave was modified and used to test the thermal characteristic of used cooking oil in the terms of temperature evolution, heating rate, and thermal efficiency. The samples were examined under various microwave power of 347W, 399W, 572W and 642W for 25 minutes of irradiation time. The char loading was tested in the level of 0, 50, and 100 g. Microwave reactor consists of microwave unit with a maximum power of 642W, a ceramic reactor, and a condenser equipped with temperature measurement system was successfully developed. It was found that microwave power and addition of absorber significantly influenced the thermal characteristic of microwave reactor. Under investigated condition, the optimum result was obtained at microwave power of 642W and 100 g of char. The condition was able to provide temperature of 480°C, heating rate of 18.2°C/min and thermal efficiency of 53% that is suitable to pyrolyze used cooking oil.

Thermal hydraulics analysis of the Advanced High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

2015-12-01

Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Thermal Reactor Safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Thermal Reactor Safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

Applying chemical engineering concepts to non-thermal plasma reactors

Science.gov (United States)

Pedro AFFONSO, NOBREGA; Alain, GAUNAND; Vandad, ROHANI; François, CAUNEAU; Laurent, FULCHERI

2018-06-01

Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas. Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge. In this work, we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes, such as laminar or plug flow, may have on the reactor performance. We do this in the particular context of the removal of pollutants by non-thermal plasmas, for which a simplified model is available. We generalise this model to different reactor configurations and, under certain hypotheses, we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime, often assumed in the non-thermal plasma literature. On the other hand, we show that a packed-bed reactor behaves very similarly to one in the plug flow regime. Beyond those results, the reader will find in this work a quick introduction to chemical reaction engineering concepts.

Thermal shield support degradation in pressurized water reactors

International Nuclear Information System (INIS)

Sweeney, F.J.; Fry, D.N.

1986-01-01

Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

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.

Thermal Energetic Reactor with High Reproduction of Fission Materials

International Nuclear Information System (INIS)

Kotov, V.M.

2012-01-01

Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

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

International Nuclear Information System (INIS)

1983-01-01

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

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

Comparison of Several Thermal Conductivity Constants for Thermal Hydraulic Calculation of Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

2017-07-01

There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

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

1979-01-01

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

A study on criticality of coupled fast-thermal core HERBE at RB reactor

Energy Technology Data Exchange (ETDEWEB)

Pesic, M; Zavaljevski, M; Milosevic, M; Stefanovic, D; Nikolic, D; Avdic, S [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia); Popovic, D; Marinkovic, P [Faculty of Electrical Engineering, Beograd (Yugoslavia)

1991-07-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

A study on criticality of coupled fast-thermal core HERBE at RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Zavaljevski, M.; Milosevic, M.; Stefanovic, D.; Nikolic, D.; Avdic, S.; Popovic, D.; Marinkovic, P.

1991-01-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

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

Passive residual energy utilization system in thermal cycles on water-cooled power reactors

International Nuclear Information System (INIS)

Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

2013-01-01

This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-30

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

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

Performance of the prism reactor's passive decay heat removal system

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

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

1983-01-01

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

Thermal baffle for fast-breeder reactor

International Nuclear Information System (INIS)

Rylatt, J.A.

1977-01-01

A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel. 3 claims, 2 figures

Determination of reactor thermal power using a more accurate method

International Nuclear Information System (INIS)

Papuga, J.; Madron, F.; Pliska, J.

2005-01-01

Reactor thermal power is an important operational parameter in many respects such as nuclear safety, reactor physics or evaluation of turbine thermal performance. Thermal power of a pressurized water reactor is determined on the basis of the steam generator thermal balance. The balance can be made in several variants differing from one another by the selection of different measuring circuits whose data are used in the balancing. In principle, no one such variant gives the true value of the thermal power. Among the variant values, the one nearest to the unknown true value of reactor thermal power is probably the value calculated with the lowest uncertainty. The determination of such uncertainty is not easy and its value can make even several percent, which has significant economic consequences. This paper presents the method of data reconciliation and its application to the data of the third of Dukovany NPP. The data reconciliation method allows to exploit all the information which process data contain. It is based on the statistical adjustment of the redundant data in such a way that the adjusted data obey generally valid laws of nature (e.g. conservation laws). Mass and energy balances based on the data not yet reconciled do not obey those laws because of measurement errors. For data reconciliation in Dukovany, a detailed model of mass and energy flows describing the 3rd unit from steam generators to alternator and condenser was set up. Laws of mass and energy conservation and phase equilibrium in water-steam systems are thus fulfilled. Moreover, the user can model momentum balances in pipelines and create other equations, which are respected during calculation. The data reconciliation is done regularly for hourly averages (Authors)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

International Nuclear Information System (INIS)

Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

2005-01-01

In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

International Nuclear Information System (INIS)

Schnitzler, Bruce G.

2012-01-01

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse (∼900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as

Thermal-hydraulic interfacing code modules for CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Thermal-hydraulic interfacing code modules for CANDU reactors

International Nuclear Information System (INIS)

Liu, W.S.; Gold, M.; Sills, H.

1997-01-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

Cross-cutting european thermal-hydraulics research for innovative nuclear systems

International Nuclear Information System (INIS)

Roelofs, F.; Class, A.; Cheng, X.; Meloni, P.; Van Tichelen, K.; Boudier, P.; Prasser, M.

2010-01-01

Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). This results in different micro- and macroscopic behavior of flow and heat transfer and requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulic issues are the subject of the 7. framework programme THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which runs from 2010 until 2014. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. (authors)

Availability of steam generator against thermal disturbance of hydrogen production system coupled to HTGR

International Nuclear Information System (INIS)

Shibata, Taiju; Nishihara, Tetsuo; Hada, Kazuhiko; Shiozawa, Shusaku

1996-01-01

One of the safety issues to couple a hydrogen production system to an HTGR is how the reactor coolability can be maintained against anticipated abnormal reduction of heat removal (thermal disturbance) of the hydrogen production system. Since such a thermal disturbance is thought to frequently occur, it is desired against the thermal disturbance to keep reactor coolability by means other than reactor scram. Also, it is thought that the development of a passive cooling system for such a thermal disturbance will be necessary from a public acceptance point of view in a future HTGR-hydrogen production system. We propose a SG as the passive cooling system which can keep the reactor coolability during a thermal disturbance of a hydrogen production system. This paper describes the proposed steam generator (SG) for the HTGR-hydrogen production system and a result of transient thermal-hydraulic analysis of the total system, showing availability of the SG against a thermal disturbance of the hydrogen production system in case of the HTTR-steam reforming hydrogen production system. (author)

A contribution for the problematic of measurements of fast-neutron-energy spectrum in thermal reactor-systems

International Nuclear Information System (INIS)

Dederichs, H.

1978-06-01

The aims of this work are to check the experimental conditions for using of a 6 Li-semiconductor-spectrometer at thermal reactor-systems and to measure the neutron-energy-spectra at the critical experiment KAHTER comparing with the theory. Using the spectrometer at thermal-neutraon-experiments questions will be attended as resolution, statistic and selection of usable nuclear data. The nuclear data will be gauged by qualified measurements, the statistic will be estimated by simulated calculations and the resolution will be improved by using the Fredholm-equation in the calculations. The calculated spectra show a good agreement with the measured spectra. Only in the energy region of maximum distribution of fission-neutrons there are little difference. The measurements show the using of the spectrometer is recommended at systems with preponderant thermal neutron-spectra, although the resolution and statistic are optimized for the spectrometer by measurements at experiments with fast neutron-spectra. (orig.) 891 RW [de

Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

International Nuclear Information System (INIS)

Novelli, A.

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Historical perspective of thermal reactor safety in light water reactors

International Nuclear Information System (INIS)

Levy, S.

1986-01-01

A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

International Nuclear Information System (INIS)

Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

2014-01-01

Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

International Nuclear Information System (INIS)

Ville Lestinen; Timo Toppila

2005-01-01

Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

International Nuclear Information System (INIS)

Shiozawa, Shusaku

2003-01-01

On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

Nuclear vapor thermal reactor propulsion technology

International Nuclear Information System (INIS)

Maya, I.; Diaz, N.J.; Dugan, E.T.; Watanabe, Y.; McClanahan, J.A.; Wen-Hsiung Tu; Carman, R.L.

1993-01-01

The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF 4 ) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF 4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (∼100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development

Benchmark testing of CENDL-2 for U-fuel thermal reactors

International Nuclear Information System (INIS)

Zhang Baocheng; Liu Guisheng; Liu Ping

1995-01-01

Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

Design of tandem mirror reactors with thermal barriers

International Nuclear Information System (INIS)

Carlson, G.A.

1980-01-01

End-plug technologies for tandem mirror reactors include high-field superconducting magnets, neutral beam injectors, and gyrotrons for electron cyclotron resonant heating (ECRH). In addition to their normal use for sustenance of the end-plug plasmas, neutral beam injectors are used for ''pumping'' trapped ions from the thermal barrier regions by charge exchange. An extra function of the axially directed pump beams is the removal of thermalized alpha particles from the reactor. The principles of tandem mirror operation with thermal barriers will be demonstrated in the upgrade of the Tandem Mirror Experiment (TMX-U) in 1981 and the tandem configuration of the Mirror fusion Test Facility (MFTF-B) in 1984

Thermal-hydraulic unreliability of passive systems

International Nuclear Information System (INIS)

Tzanos, C.P.; Saltos, N.T.

1995-01-01

Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

Thermal performance and efficiency of supercritical nuclear reactors

International Nuclear Information System (INIS)

Romney Duffey; Tracy Zhou; Hussam Khartabil

2009-01-01

The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)

Pellet bed reactor for nuclear thermal propelled vehicles

International Nuclear Information System (INIS)

El-Genk, M.; Morley, N.J.; Haloulakos, V.E.

1991-01-01

The Pellet Bed Reactor (PeBR) concept is capable of operating at a high power density of up to 3.0 kWt/cu cm and an exit hydrogen gas temperature of 3000 K. The nominal reactor thermal power is 1500 MW and the reactor core is 0.80 m in diameter and 1.3 m high. The nominal PeBR engine generates a thrust of approximately 315 kN at a specific impulse of 1000 s for a mission duration to Mars of 250 days requiring a total firing time of 170 minutes. Because of its low diameter-to-height ratio, PeBR has enough surface area for passive removal of the decay heat from the reactor core. The reactor is equipped with two independent shutdown mechanisms; 8-B4C safety rods and 26 BeO/B4C control drums; each system is capable of operating and scraming the reactor safely. Due to the absence of core internal support structures, the PeBR can be fueled and refueled in orbit using the vacuum of space. These unique features of the PeBR provide for safety during launch, simplicity of handling, deployment, and end-of-life disposal, and vehicle extended lifetime. 11 refs

A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

1998-03-01

The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

Real time thermal hydraulic model for high temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

2013-01-01

A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

Yamada, K.; Aksan, S. N.

2012-01-01

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

Application of the REMIX thermal mixing calculation program for the Loviisa reactor

International Nuclear Information System (INIS)

Kokkonen, I.; Tuomisto, H.

1987-08-01

The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

Thermal reactor benchmark testing of 69 group library

International Nuclear Information System (INIS)

Liu Guisheng; Wang Yaoqing; Liu Ping; Zhang Baocheng

1994-01-01

Using a code system NSLINK, AMPX master library in WIMS 69 groups structure are made from nuclides relating to 4 newest evaluated nuclear data libraries. Some integrals of 10 thermal reactor benchmark assemblies recommended by the U.S. CSEWG are calculated using rectified PASC-1 code system and compared with foreign results, the authors results are in good agreement with others. 69 group libraries of evaluated data bases in TPFAP interface file are generated with NJOY code system. The k ∞ values of 6 cell lattice assemblies are calculated by the code CBM. The calculated results are analysed and compared

Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

International Nuclear Information System (INIS)

Marsodi; Mulyanto; Kitamoto, Asashi.

1994-01-01

Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

Thermal hydraulic simulation of the CANDU nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

2017-07-01

The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

2017-06-15

Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

2009-01-01

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

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.

Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

International Nuclear Information System (INIS)

Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia

2017-01-01

MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

Energy Technology Data Exchange (ETDEWEB)

Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

2017-07-01

MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

Utilization of thorium in thermal reactors

International Nuclear Information System (INIS)

Srinivasan, K.R.; Nakra, A.N.

1978-01-01

Large deposits of thorium are found in India. 233 U produced by neutron capture in 232 Th is a more valuable fuel for thermal reactors than the plutonium that results from capture in 238 U. These two facts are the main reasons for the interest in utilizing thorium in power reactors. But natural thorium does not contain any fissile material and its capture cross section is nearly two and a half times that of 238 U. These have made the fuelling cost high. However, in certain conditions and certain types of reactors the costs are comparable with those using uranium fuel. The relative cost effectiveness of different fuels is discussed. Apart from long term interest, the short term interest of using thorium fuel in RAPP type reactors is also briefly described. Finally the reactor physics experiments using thorium fuel and their comparison with calculations are presented. (author)

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

maximum neutron flux at thermal nuclear reactors

International Nuclear Information System (INIS)

Strugar, P.

1968-10-01

Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

An overview of the potential of the CANDU reactor as a thermal breeder

International Nuclear Information System (INIS)

Slater, J.B.

1977-02-01

This paper is concerned with the use of thorium as a fuel in the existing CANDU concept. The neutron balance of the reactor core is analyzed and an assessment is made of the potential for development of a thermal 'breeder' reactor system. It is concluded that while the SSET cycle (i.e. self-sufficient equilibrium thorium cycle) appears feasible, there is little potential for developing a significant 'breeding' fuel cycle if current reactor operating capability and capital costs are to be maintained. (author)

Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

International Nuclear Information System (INIS)

1980-01-01

Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

International Nuclear Information System (INIS)

Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

2009-01-01

This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

2010-07-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

Computer code system for the R and D of nuclear fuel cycle with fast reactor. 2. Development and application of analytical evaluation system for thermal striping phenomena

International Nuclear Information System (INIS)

Muramatsu, Toshiharu

2001-01-01

Fluid-structure thermal interaction phenomena characterized by stationary random temperature fluctuations, namely thermal striping are observed in the downstream region such as a T-junction piping system of liquid metal fast reactors (LMFRs). Therefore, the piping wall located in the downstream region must be protected against the stationary random thermal process, which might induce high-cycle fatigue. This paper describes the evaluation system based on numerical simulation methods consisting of three thermohydraulics computer programs AQUA, DINUS-3 and THEMIS and of three thermomechanical computer programs BEMSET, FINAS and CANIS, for the thermal striping developed at Japan Nuclear Cycle Development Institute (JNC). Verification results for each computer code and the system are also introduced based on out-of-pile experimental data using water and sodium as working fluids. (author)

Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

Energy Technology Data Exchange (ETDEWEB)

Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

A new method for evaluation and correction of thermal reactor power and present operational applications

International Nuclear Information System (INIS)

Langenstein, M.; Streit, S.; Laipple, B.; Eitschberger, H.

2005-01-01

describes the quality control of important process parameter. Applied to the thermal reactor power, the statistical certainty of warranting the allowable value can be quantified. This quantification allows keeping a safety margin in agreement with the authority. This paper presents the operational application of this method at an operating plant and describes the additional use of process data reconciliation for acceptance tests, system and component diagnosis. (authors)

Transient thermal characteristics of a core channel in a molten salt reactor

International Nuclear Information System (INIS)

Sakashita, H.; Ishiguro, R.; Sugiyama, K.

1987-01-01

The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

International Nuclear Information System (INIS)

Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.

1997-07-01

The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: 1) basic requirement for the integral reactor system design 2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system 3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs

Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.

1997-07-01

The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: (1) basic requirement for the integral reactor system design (2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system (3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

Energy Technology Data Exchange (ETDEWEB)

O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

REACTOR GROUT THERMAL PROPERTIES

Energy Technology Data Exchange (ETDEWEB)

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

Thermal aspects of a superconducting coil for fusion reactor

International Nuclear Information System (INIS)

Yeh, H.T.

1975-01-01

Computer models are used to simulate both localized and extensive thermal excursions in a large superconducting magnet for fusion reactor. Conditions for the failure of fusion magnet due to thermal excursion are delineated. Designs to protect the magnet against such thermal excursion are evaluated

Laser fusion power reactor system (LFPRS)

International Nuclear Information System (INIS)

Kovacik, W.P.

1977-01-01

This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

Thermal Energetic Reactor with High Reproduction of Fission Materials

Directory of Open Access Journals (Sweden)

Vladimir M. Kotov

2012-01-01

On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

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

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

International Nuclear Information System (INIS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-01-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed

Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

International Nuclear Information System (INIS)

1991-01-01

Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

Science.gov (United States)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

International Nuclear Information System (INIS)

Ramani, D.T.

1977-01-01

The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

N Reactor thermal plume characterization during Pu-only mode of operation

Energy Technology Data Exchange (ETDEWEB)

Ecker, R.M.; Thompson, F.L.; Whelan, G.

1983-04-01

Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Energy Technology Data Exchange (ETDEWEB)

Neil Todreas; Pavel Hejzlar

2008-06-30

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations

International Nuclear Information System (INIS)

Robinson, G.S.

1985-08-01

The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent

Thermal performance of Egypt's research reactor core (ET-RR-1)

International Nuclear Information System (INIS)

Khattab, M.; Mariy, A.

1986-01-01

The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

Energy Technology Data Exchange (ETDEWEB)

Abou-El-Maaty, Talal [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)], E-mail: talal22969@yahoo.com; Abdelhady, Amr [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)

2009-03-15

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction ({beta}{sub eff}) and the prompt neutron generation time ({lambda}). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both {beta}{sub eff} and {lambda}.

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Efficient cycles for carbon capture CLC power plants based on thermally balanced redox reactors

KAUST Repository

Iloeje, Chukwunwike

2015-10-01

© 2015 Elsevier Ltd. The rotary reactor differs from most alternative chemical looping combustion (CLC) reactor designs because it maintains near-thermal equilibrium between the two stages of the redox process by thermally coupling channels undergoing oxidation and reduction. An earlier study showed that this thermal coupling between the oxidation and reduction reactors increases the efficiency by up to 2% points when implemented in a regenerative Brayton cycle. The present study extends this analysis to alternative CLC cycles with the objective of identifying optimal configurations and design tradeoffs. Results show that the increased efficiency from reactor thermal coupling applies only to cycles that are capable of exploiting the increased availability in the reduction reactor exhaust. Thus, in addition to the regenerative cycle, the combined CLC cycle and the combined-regenerative CLC cycle are suitable for integration with the rotary reactor. Parametric studies are used to compare the sensitivity of the different cycle efficiencies to parameters like pressure ratio, turbine inlet temperature, carrier-gas fraction and purge steam generation. One of the key conclusions from this analysis is that while the optimal efficiency for regenerative CLC cycle was the highest of the three (56% at 3. bars, 1200. °C), the combined-regenerative cycle offers a trade-off that combines a reasonably high efficiency (about 54% at 12. bars, 1200. °C) with much lower gas volumetric flow rate and consequently, smaller reactor size. Unlike the other two cycles, the optimal compressor pressure ratio for the regenerative cycle is weakly dependent on the design turbine inlet temperature. For the regenerative and combined regenerative cycles, steam production in the regenerator below 2× fuel flow rate improves exhaust recovery and consequently, the overall system efficiency. Also, given that the fuel side regenerator flow is unbalanced, it is more efficient to generate steam from the

Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

International Nuclear Information System (INIS)

Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

2002-01-01

KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

The promise and challenges of cermet fueled nuclear thermal propulsion reactors

International Nuclear Information System (INIS)

Brengle, R.G.; Harty, R.B.; Bhattacharyya, S.K.

1993-06-01

The use of cermet fuels in nuclear thermal propulsion systems was examined and the characteristics of systems using these fuel forms is discussed in terms of current mission and safety requirements. For use at high temperatures cermet fueled reactors utilize ceramic fuels with refractory metals as the matrix material. Cermet fueled reactors tend to be heavy when compared to concepts that utilize graphite as the fuel matrix because of the high density of the refractory metal matrix which makes up 20-40 percent of the total volume. On the positive side the metal matrix is strong and more resistant to loads from either the launch or flow induced vibration. The compatibility of the tungsten cermet with hydrogen is excellent and lifetimes of several hours is certainly achievable. Probably the biggest drawback to cermet nuclear thermal propulsion concepts is that the amount of actual data to support the theoretical conclusions is small. In fact there is no data under representative conditions of temperature, propellant and flux for the required fuel burnup. Although cermet systems appear to be attractive, the lack of fuel data at representative conditions does not allow reliable comparisons of cermet systems to systems where fuel data is available. 10 refs

An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

Energy Technology Data Exchange (ETDEWEB)

Haihua Zhao; Per F. Peterson

2010-10-01

Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

International Nuclear Information System (INIS)

Monteleone, S.

1994-04-01

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

Thermal properties of reactors and some instabilities

Energy Technology Data Exchange (ETDEWEB)

Hearfield, F.

1979-03-01

A discussion covers the thermal properties of adiabatic reactors and the failure of the reaction rate to increase with increasing temperature due to depletion of reagents, transition to mass transfer control, or reduction of adsorption at catalytic surfaces; non-adiabatic reactors and factors upsetting the balance between heat generation and removal and possibly causing a runaway reaction, including loss of agitation loop circulation, and cooling or heating media; multiple steady states, i.e. multiple balances between heat generation and removal, for a continuous stirred tank reactor and the conditions necessary for stability of a steady state; and the temperature distribution in a tubular reactor, including mechanisms for feedback of heat from downstream to upstream in the reactor, e.g. heat conduction and radiation from hot catalyst, or an added heat exchanger. Three case histories are presented in which reactants accumulated in the reactors and cooling was decreased, permitting the occurrence of violent runaway reactions.

C.E.C. - cod for calculus of the evolution fuel for thermal reactors

International Nuclear Information System (INIS)

Biciolla, L.; Marcu, G.; Mociornita, G.

1975-01-01

The study of ''burnup'' into thermal reactor involves two main aspects: the economic one and another regarding the reactor operation, its stability and control. In the CEC-code written in FORTRAN IV language was analysed the change of the isotopic composition of nuclear fuel from thermal reactor during its operation

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

Sensitivity Analysis of Reactor Regulating System for SMART

International Nuclear Information System (INIS)

Jeon, Yu Lim; Kang, Han Ok; Lee, Seong Wook; Park, Cheon Tae

2009-01-01

The integral reactor technology is one of the Small and Medium sized Reactor (SMR) which has recently come into a spotlight due to its suitability for various fields. SMART (System integrated Modular Advanced ReacTor), a small sized integral type PWR with a rated thermal power of 330MWt is one of the advanced SMR. SMART developed by the Korea Atomic Energy Research Institute (KAERI), has a capacity to provide 40,000 m3 per day of potable water and 90 MW of electricity (Chang et al., 2000). Figure 1 shows the SMART which adopts a sensible mixture of new innovative design features and proven technologies aimed at achieving highly enhanced safety and improved economics. Design features contributing to a safety enhancement are basically inherent safety improving features and passive safety features. Fundamental thermal-hydraulic experiments were carried out during the design concepts development to assure the fundamental behavior of major concepts of the SMART systems. A TASS/SMR is a suitable code for accident and performance analyses of SMART. In this paper, we proposed a new power control logic for stable operating outputs of Reactor Regulating System (RRS) of SMART. We analyzed the sensitivity of operating parameter for various operating conditions

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

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

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

Release of WIMS10: a versatile reactor physics code for thermal and fast systems - 15467

International Nuclear Information System (INIS)

Lindley, B.A.; Newton, T.D.; Hosking, J.G.; Smith, P.N.; Powney, D.J.; Tollit, B.; Smith, P.J.

2015-01-01

the WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations using a range of flux solutions methods to produce the neutronic libraries for use in PANTHER or other whole core analysis codes. With the release of WIMS10, the range of problems which WIMS can solve has been greatly extended. A WIMS/PANTHER calculation route has been developed and validated for part MOX-fuelled PWRs, with calculations showing excellent agreement with 2D core deterministic and Monte Carlo transport solutions. A flexible geometry 3D method of characteristics transport solver, CACTUS3D has been added to the code. CACTUS3D has been benchmarked for a 3D BWR assembly model, and was in good agreement with a direct 172-group solution in the Monte Carlo code MONK. Fast reactor calculations using the ECCO deterministic calculation route have been validated using experimental data from the ZEBRA reactor. Power deposition can be treated through following neutrons and/or photons to their point of interaction. The improved methodology is shown to give more accurate calculation of heat deposition and improve agreement between calculated and measured detector responses for part MOX-fuelled cores. (authors)

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

Preliminary thermal sizing of intermediate heat exchanger for NHDD system

International Nuclear Information System (INIS)

Kim, Chan Soo; Hong, Sung Deok; Kim, Yong Wan; Chang, Jongh Wa

2009-01-01

Nuclear Hydrogen Development and Demonstration (NHDD) system is a Very High Temperature gascooled Reactor (VHTR) coupled with hydrogen production systems. Intermediate heat exchanger transfers heat from the nuclear reactor to the hydrogen production system. This study presented the sensitivity analysis on a preliminary thermal sizing of the intermediate heat exchanger. Printed Circuit Heat Exchanger (PCHE) was selected for the thermal sizing because the printed circuit heat exchanger has the largest compactness among the heat exchanger types. The analysis was performed to estimate the effect of key parameters including the operating condition of the intermediate system, the geometrical factors of the PCHE, and the working fluid of the intermediate system.

The analysis of thermal-hydraulic performances of nuclear ship reactor

International Nuclear Information System (INIS)

Wakabayashi, Shinshichi; Hamada, Masao

1975-01-01

Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

Density dependence of reactor performance with thermal confinement scalings

International Nuclear Information System (INIS)

Stotler, D.P.

1992-03-01

Energy confinement scalings for the thermal component of the plasma published thus far have a different dependence on plasma density and input power than do scalings for the total plasma energy. With such thermal scalings, reactor performance (measured by Q, the ratio of the fusion power to the sum of the ohmic and auxiliary input powers) worsens with increasing density. This dependence is the opposite of that found using scalings based on the total plasma energy, indicating that reactor operation concepts may need to be altered if this density dependence is confirmed in future research

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.

Simulation of Thermal-hydraulic Process in Reactor of HTR-PM

International Nuclear Information System (INIS)

Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

2014-01-01

This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

Development of a system model for advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

International Nuclear Information System (INIS)

Schweizer, Fernando Lage Araújo

2014-01-01

The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

Determining space-energy distribution of thermal neutrons in heterogeneous cylindrically symmetric reactor cell, Master Thesis

International Nuclear Information System (INIS)

Matausek, M. V.

1966-06-01

A combination of multigroup method and P 3 approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)

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

International Nuclear Information System (INIS)

Sekimizu, Koichi

1982-01-01

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

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

Maximum neutron flux in thermal reactors; Maksimum neutronskog fluksa kod termalnih reaktora

Energy Technology Data Exchange (ETDEWEB)

Strugar, P V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1968-07-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples.

Investigation on in-vessel thermal transients in a fast breeder reactor

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Kasahara, Naoto

1999-01-01

Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

Development potential for thermal reactors and their fuel cycles

International Nuclear Information System (INIS)

Rogers, J.T.; Dodds, H.L. Jr.; Florido, P.C.; Gat, U.; Kondo, S.; Spinks, N.S.

1997-01-01

Water-cooled reactors represent the only types which have reached widespread commercial use up to the present day. Given the plentiful supply of uranium in the world today, this situation might be expected to continue for some time into the future. Nevertheless, for different reasons several countries consider that either new reactor types should be developed or that existing types should be improved substantially. The predominant reason in the short term is to improve the competitive position of nuclear energy supply versus fossil energy. In the longer term, regional and national fuel supply independence may become the dominant driving forces. This paper outlines several possible means for responding to these driving forces. It is not meant to include an exhaustive list of all possibilities, but only to illustrate some alternative routes. These routes range from enhancement of existing reactor concepts to combination of nuclear with fossil systems, and finally to the introduction of radically new thermal reactor concepts. Each of these has its obvious advantages and disadvantages and will come forward or will recede depending on technical feasibility, economics, long-term sustainability, and national policy. (author)

Passive cooling system for nuclear reactor containment structure

Science.gov (United States)

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Isotopes accumulation in the thermal column of TRIGA reactor

International Nuclear Information System (INIS)

Iorgulis, C.; Diaconu, D.; Gugiu, D.; Csaba, R.

2013-01-01

The correlation of impurity observed in the virgin graphite and radionuclide content and activities measured in the irradiated graphite needs to know the irradiated history. This is a challenging process if impurity content and irradiation conditions are not accurately known. This is the case of the irradiated graphite in the thermal column of Institute for Nuclear Research Pitesti (INR)14 MW TRIGA reactor. To overcome incomplete impurity content and the unknown position in the column of the measured irradiated graphite available for characterisation and comparison, a set of preliminary simulations were performed. Following Eu 152 /Eu 154 ration they allowed the estimation of an impurity content and irradiation conditions leading to measured activities. Based on these data the radio-isotope accumulation in different positions in the thermal column was predicted. Modelling performed by INR used advanced prediction packages (e.g. WIMS, MCNP ORIGEN-S from Scale 5) to assess the isotopic content of MTR graphite types with irradiation history specific for a TRIGA research reactor. Some certain calculations points from the column were selected in order to model the burnup and isotopes productions using ORIGEN from SCALE code system. (authors)

Parametric study on thermal-hydraulic characteristics of high conversion light water reactor

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.

1988-11-01

To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)

Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

International Nuclear Information System (INIS)

Meuschke, R.E.; Boyd, C.H.

1989-01-01

This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

Assessing thermal conductivity of composting reactor with attention on varying thermal resistance between compost and the inner surface.

Science.gov (United States)

Wang, Yongjiang; Niu, Wenjuan; Ai, Ping

2016-12-01

Dynamic estimation of heat transfer through composting reactor wall was crucial for insulating design and maintaining a sanitary temperature. A model, incorporating conductive, convective and radiative heat transfer mechanisms, was developed in this paper to provide thermal resistance calculations for composting reactor wall. The mechanism of thermal transfer from compost to inner surface of structural layer, as a first step of heat loss, was important for improving insulation performance, which was divided into conduction and convection and discussed specifically in this study. It was found decreasing conductive resistance was responsible for the drop of insulation between compost and reactor wall. Increasing compost porosity or manufacturing a curved surface, decreasing the contact area of compost and the reactor wall, might improve the insulation performance. Upon modeling of heat transfers from compost to ambient environment, the study yielded a condensed and simplified model that could be used to conduct thermal resistance analysis for composting reactor. With theoretical derivations and a case application, the model was applicable for both dynamic estimation and typical composting scenario. Copyright © 2016 Elsevier Ltd. All rights reserved.

Thermal-hydraulic modeling needs for passive reactors

Energy Technology Data Exchange (ETDEWEB)

Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

Thermal-hydraulic modeling needs for passive reactors

International Nuclear Information System (INIS)

Kelly, J.M.

1997-01-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken

Phenomena occuring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1990-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

A hybrid solar and chemical looping combustion system for solar thermal energy storage

International Nuclear Information System (INIS)

Jafarian, Mehdi; Arjomandi, Maziar; Nathan, Graham J.

2013-01-01

Highlights: ► A novel solar–CLC hybrid system is proposed which integrates a CLC with solar thermal energy. ► The oxygen carrier particles are used as storage medium for thermal energy storage. ► A solar cavity reactor is proposed for fuel reactor. ► The absorbed solar energy is stored in the particles to produce a base heat load. -- Abstract: A novel hybrid of a solar thermal energy and a chemical looping combustion (CLC) system is proposed here, which employs the oxygen carrier particles in a CLC system to provide diurnal thermal energy storage for concentrated solar thermal energy. In taking advantage of the chemical and sensible energy storage systems that are an inherent part of a CLC system, this hybrid offers potential to achieve cost effective, base load power generation for solar energy. In the proposed system, three reservoirs have been added to a conventional CLC system to allow storage of the oxygen carrier particles, while a cavity solar receiver has been chosen for the fuel reactor. The performance of the system is evaluated using ASPEN PLUS software, with the model being validated using independent simulation result reported previously. Operating temperature, solar efficiency, solar fraction, exergy efficiency and the fraction of the solar thermal energy stored for a based load power generation application are reported.

Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

Energy Technology Data Exchange (ETDEWEB)

Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

2017-04-15

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Passive Decay Heat Removal System for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

2015-10-15

Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C.

2011-01-01

The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

-isotropic) particles. They also use graphite as moderator. The results of the thermal analysis obtained in this work demonstrated the ability of the RELAP5-3D code to reproduce the behavior of the simulated core reactors. Thus, this study adds knowledge to the several researches that have been carried out on the thermal hydraulic analysis of these new systems, searching for models capable of reproducing their thermal behavior, especially in cases of transient situations or accident. This thesis present new studies, especially detailed investigation on the heat transfer across the fuel. (author)

Conceptual research on reactor core physics for accelerator driven sub-critical reactor

International Nuclear Information System (INIS)

Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

2000-01-01

The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

Analytical prediction and experimental verification of reactor safety system injection transient

International Nuclear Information System (INIS)

Roy, B.N.; Nomm, E.

1991-01-01

This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs

Thermal barrier and support for nuclear reactor fuel core

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-01-01

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

Thermal stratification of sodium in the BN 600 reactor

International Nuclear Information System (INIS)

Obmelukhin, J.A.; Obukhov, P.I.; Rinejskij, A.A.; Sobolev, V.A.; Sherbakov, S.I.

1983-01-01

The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)

Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

International Nuclear Information System (INIS)

Santos Bastos, W. dos

1995-01-01

These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

Energy Technology Data Exchange (ETDEWEB)

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

In-place thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1985-04-01

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

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

Development of hot water supply system for a small district heating reactor

International Nuclear Information System (INIS)

Murase, Toshihiko; Narabayashi, Tadashi; Shimazu, Yoichiro

2007-01-01

On the earth, there are many environmental problems. For example, rapid increase of world population causes the enormous consumption of fossil fuel and emission of CO 2 into the global air. Now, mankaind faced to deal with these serious problems. One solution for these problems is utilization of nuclear reactors. Currently, about 65% of thermal output of a nuclear reactor is thrown away to the sea or the atmosphere through a turbine condenser. When a hot-water pipeline from a nuclear plant will be constructed, the exhaust heat from nuclear reactor will able to be utilized. Therefore, authors began to study nuclear power plant system for district heating. This reactor is based on a PWR plant. Its thermal output is 10 MWth and its electrical output is 3.4 MW. The nuclear plant supply electricity and heat for 2000 to 3000 houses. The plant aim to supply all the energy for the adjacent pepole's life, for example, heat, electricity and hydrogen for fuel battery car. This total-energy supply system assumed to be built in Northern area such as Hokkaido in Japan. In order to develop an optimum thermal design method for the system, heat transport experiments and thermal-hydraulic calculations were carried out. Using a metal pipe covered with foam-polyurethane thermal insulator, feed-water temperature and return-water temperature was measured to evaluate heat loss. As the result, the heat loss from the hot-water temperature was very little. The thermal-hydraulic calculation method was verified and applied to actual pipeline size calculation. The result of heat loss calculation will be 0.2degC/5 km. considering these results, the best pipe specification was obtained. (author)

Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

International Nuclear Information System (INIS)

Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

2012-06-01

The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

International Nuclear Information System (INIS)

Nassersharif, B.; Peer, J.S.; DeHart, M.D.

1987-01-01

Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

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.

Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Ustun, G.; Durmayaz, A.

2002-01-01

Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

Novel “open-sorption pipe” reactor for solar thermal energy storage

International Nuclear Information System (INIS)

Aydin, Devrim; Casey, Sean P.; Chen, Xiangjie; Riffat, Saffa

2016-01-01

Highlights: • A novel ‘open sorption pipe’ heat storage was experimentally investigated. • Effect of absolute moisture levels on heat storage performance was analyzed. • Hygrothermal-cyclic performances of Zeolite 13X and vermiculite–calcium chloride were compared. • Vermiculite–calcium chloride has more durable performance than Zeolite at 80 °C regeneration temperature. • Sorption pipe system using vermiculite–calcium chloride provides energy storage density of 290 kW h/m"3. - Abstract: In the last decade sorption heat storage systems are gaining attention due to their high energy storage density and long term heat storage potential. Sorption reactor development is vital for future progress of these systems however little has done on this topic. In this study, a novel sorption pipe reactor for solar thermal energy storage is developed and experimentally investigated to fulfill this gap. The modular heat storage system consists of sorption pipe units with an internal perforated diffuser pipe network and the sorption material filled in between. Vermiculite–calcium chloride composite material was employed as the sorbent in the reactor and its thermal performance was investigated under different inlet air humidity levels. It was found that, a fourfold increase of absolute humidity difference of air led to approximately 2.3 times boost in average power output from 313 W to 730 W and an 8.8 times boost of average exergy from 4.8 W to 42.3 W. According to the testing results, each of three sorption pipes can provide an average air temperature lift of 24.1 °C over 20 h corresponding to a system total energy storage capacity of 25.5 kW h and energy storage density of 290 kW h/m"3. Within the study, vermiculite–calcium chloride performance was also compared with the widely investigated Zeolite 13X. Vermiculite–calcium chloride showed a good cyclic ability at regeneration temperature of 80 °C with a steadier thermal performance than Zeolite

Methane-steam reforming by molten salt - membrane reactor using concentrated solar thermal energy

International Nuclear Information System (INIS)

Watanuki, K.; Nakajima, H.; Hasegawa, N.; Kaneko, H.; Tamaura, Y.

2006-01-01

By utilization of concentrated solar thermal energy for steam reforming of natural gas, which is an endothermic reaction, the chemical energy of natural gas can be up-graded. The chemical system for steam reforming of natural gas with concentrated solar thermal energy was studied to produce hydrogen by using the thermal storage with molten salt and the membrane reactor. The original steam reforming module with hydrogen permeable palladium membrane was developed and fabricated. Steam reforming of methane proceeded with the original module with palladium membrane below the decomposition temperature of molten salt (around 870 K). (authors)

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

2005-07-01

Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

A thermal packed-bed reactor and a silent discharge plasma cell for a two-stage treatment system

International Nuclear Information System (INIS)

Godoy-Cabrera, O G; Lopez-Callejas, R; Mercado-Cabrera, A; Barocio, S R; Valencia, R; Munoz-Castro, A; Eguiluz, R Pena; Piedad-Beneitez, A de la

2006-01-01

Dielectric barrier discharge cells (DBDCs) have proved their efficiency in the generation of cold plasmas for hazardous organic compound degradation. Here, we describe the design and construction of a dual thermal packed-bed reactor and DBDC-based system to carry out the degradation of hazardous organic compounds in both liquid and gas phases. The main components of this system are: (i) the thermal treatment system (ii) DBDC and (iii) resonant inverters of low (3.3 kHz) and high (100 kHz) calculated frequencies. The definition of the cell physical parameters considers: (a) a first-order degradation ratio of the compound and (b) the air breakdown at atmospheric pressure as a function of the transport carrier gas. The power consumed by the cells during the discharges was computed theoretically and experimentally. Using the dual system along with a gas chromatography diagnostic system, highly efficient degradations of a test compound (benzene) have been obtained, reaching 99.950% in the case of a cell experimentally operated at 3.3 kHz and up to 99.996% in another one at 94.3 kHz. An additional 3.7 times reduction in the latter case residence time with respect to the low frequency cell has been found

Thermal insulation of the high-temperature helium-cooled reactors

International Nuclear Information System (INIS)

Kharlamov, A.G.; Grebennik, V.N.

1979-01-01

Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper

Review of the nuclear reactor thermal hydraulic research in ocean motions

Energy Technology Data Exchange (ETDEWEB)

Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn

2017-03-15

The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

Review of the nuclear reactor thermal hydraulic research in ocean motions

International Nuclear Information System (INIS)

Yan, B.H.

2017-01-01

The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca

International Nuclear Information System (INIS)

Stefanovic, D.; Pesic, M.

1995-01-01

A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

International Nuclear Information System (INIS)

Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

International Nuclear Information System (INIS)

Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

2015-01-01

A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

Measuring technique of super high temperature thermal properties of reactor core materials

International Nuclear Information System (INIS)

Ono, Akira; Baba, Tetsuya; Watanabe, Hideo; Matsumoto, Tsuyoshi

1998-01-01

In this study, thermal properties of reactor core materials used for water cooled reactors and FBR were tried to develop a technique to measure their melt states at less than 3,000degC in order to contribute more correct evaluation of the reactor core behavior at severe accident. Then, a thermal property measuring method of high temperature melt by using floating method was investigated and its fundamental design was begun to investigate under a base of optimum judgement on the air flow floating throw-down method. And, in order to measure emissivity of melt specimen surface essential for correct temperature measurement using the throw down method, a spectroscopic emissivity measuring unit using an ellipsometer was prepared and induced. On the thermal properties measurement using the holding method, a specimen container to measure thermal diffusiveness of the high temperature melts by using laser flashing method was tried to prepare. (G.K.)

Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis

Energy Technology Data Exchange (ETDEWEB)

Kadioglu, Samet Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)’s thermal-fluids code) built on top of an other INL’s product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.

A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis

International Nuclear Information System (INIS)

Kadioglu, Samet Y.; Berry, Ray; Martineau, Richard

2016-01-01

A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)'s thermal-fluids code) built on top of an other INL's product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.

Systems aspects of a space nuclear reactor power system

International Nuclear Information System (INIS)

Jaffe, L.; Fujita, T.; Beatty, R.

1988-01-01

Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: Power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, attitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly

Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

Thermal-hydraulic development a small, simplified, proliferation-resistant reactor

International Nuclear Information System (INIS)

Farmer, M. T.; Hill, D. J.; Sienicki, J. J.; Spencer, B. W.; Wade, D. C.

1999-01-01

This paper addresses thermal-hydraulics related criteria and preliminary concepts for a small (300 MWt), proliferation-resistant, liquid-metal-cooled reactor system. A main objective is to assess what extent of simplification is achievable in the concepts with the primary purpose of regaining economic competitiveness. The approach investigated features lead-bismuth eutectic (LBE) and a low power density core for ultra-long core lifetime (goal 15 years) with cartridge core replacement at end of life. This potentially introduces extensive simplifications resulting in capital cost and operating cost savings including: (1) compact, modular, pool-type configuration for factory fabrication, (2) 100+% natural circulation heat transport with the possibility of eliminating the main coolant pumps, (3) steam generator modules immersed directly in the primary coolant pool for elimination of the intermediate heat transport system, and (4) elimination of on-site fuel handling and storage provisions including rotating plug. Stage 1 natural circulation model and results are presented. Results suggest that 100+% natural circulation heat transport is readily achievable using LBE coolant and the long-life cartridge core approach; moreover, it is achievable in a compact pool configuration considerably smaller than PRISM A (for overland transportability) and with peak cladding temperature within the existing database range for ferritic steel with oxide layer surface passivation. Stage 2 analysis follows iteration with core designers. Other thermal hydraulic investigations are underway addressing passive, auxiliary heat removal by air cooling of the reactor vessel and the effects of steam generator tube rupture

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1977-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1978-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

Nuclear reactor plants and control systems therefor

International Nuclear Information System (INIS)

de Boer, G.A.; de Hex, M.

1976-01-01

A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

History of the 185-/189-D thermal hydraulics laboratory and its effects on reactor operations at the Hanford Site

International Nuclear Information System (INIS)

Gerber, M.S.

1994-09-01

The 185-D deaeration building and the 189-D refrigeration building were constructed at Hanford during 1943 and 1944. Both buildings were constructed as part of the influent water cooling system for D reactor. The CMS studies eliminated the need for 185-D function. Early gains in knowledge ended the original function of the 189-D building mission. In 1951, 185-D and 189-D were converted to a thermal-hydraulic laboratory. The experiments held in the thermal-hydraulic lab lead to historic changes in Hanford reactor operations. In late 1951, the exponential physics experiments were moved to the 189-D building. In 1958, new production reactor experiments were begun in 185/189-D. In 1959, Plutonium Recycle Test Reactor experiments were added to the 185/189-D facility. By 1960, the 185/189-D thermal hydraulics laboratory was one of the few full service facilities of its type in the nation. During the years 1961--1963 tests continued in the facility in support of existing reactors, new production reactors, and the Plutonium Recycle Test Reactor. In 1969, Fast Flux Test Facility developmental testings began in the facility. Simulations in 185/189-D building aided in the N Reactor repairs in the 1980's. In 1994 the facility was nominated to the National Register of Historic Places, because of its pioneering role over many years in thermal hydraulics, flow studies, heat transfer, and other reactor coolant support work. During 1994 and 1995 it was demolished in the largest decontamination and decommissioning project thus far in Hanford Site history

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1981-01-01

In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)

Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

International Nuclear Information System (INIS)

Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

1996-01-01

Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

1994-01-01

Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

International Nuclear Information System (INIS)

Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

2012-01-01

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

Energy Technology Data Exchange (ETDEWEB)

Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

2013-07-01

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

International Nuclear Information System (INIS)

Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.

2013-01-01

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

International Nuclear Information System (INIS)

Asaoka, Takumi; Watanabe, N.

1976-01-01

In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

International Nuclear Information System (INIS)

Sobajima, Makoto

1985-08-01

In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

Research Reactor Power Control System Design by MATLAB/SIMULINK

International Nuclear Information System (INIS)

Baang, Dane; Suh, Yong Suk; Kim, Young Ki; Im, Ki Hong

2013-01-01

In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, B. D; Kim, K. D.

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

International Nuclear Information System (INIS)

Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

1992-01-01

The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

A cermet fuel reactor for nuclear thermal propulsion

International Nuclear Information System (INIS)

Kruger, G.

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Thermally-insulating layer for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The thermally-insulating layer has been designed both for insulating surfaces within the core of a nuclear reactor and transmitting loads such as the core-weight. Said layer comprises a layer of bricks and a layer of tiles with smaller clearance between the tiles than between the bricks, the latter having a reduced cross-section against the tiles so as to be surrounded by relatively large interconnected ducts forming a continuous chamber behind the tile-layer in order to induce a substantial decreases in the transverse flow of the reactor-core coolant. The core preferably comprises hexagonal columns supported by rhomb-shaped plates, with channels distributed so as to mix the coolant of twelve columns. The plates are separated from support-tiles by means of pillars [fr

COOLOD, Steady-State Thermal Hydraulics of Research Reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-01-01

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

Design of a molten heavy-metal coolant and target for fast-thermal accelerator driven sub-critical system (ADS)

International Nuclear Information System (INIS)

Satyamurthy, P.; Degwekar, S.B.; Nema, P.K.

2001-01-01

Accelerator Driven sub-critical Systems (ADS) have evoked considerable interest in recent years. The Energy Amplifier concept developed by C. Rubbia and others at CERN incorporates a buoyancy driven, lead-coolant primary system for extracting the heat generated in the fast reactor as well as that in neutron spallation target. In earlier publications, our BARC group has proposed a one-way coupled booster reactor system which could be operated at proton beam currents as low as 1-2 mA for a power output of 750 MW th . Here, the basic idea is to have a fast booster reactor zone of low power (- 100 MW th ) which is separated by a large gap from the main thermal reactor zone. In this arrangement, the spallation neutron source feeds neutrons to the fast reactor zone where neutrons are further multiplied. Further in this system, the neutrons from the booster region enter the main reactor but very few neutrons from main reactor return to booster, thus ensuring one-way coupling. In earlier work, several possible configurations of the booster and thermal regions were presented. In the present work, we describe an engineering design particularly with respect to thermal hydraulics of lead/lead-bismuth eutectic coolant also acting as spallation neutron source. This hybrid ADS reactor consists of fast and thermal reactor zones producing about 100 MW th and 650 MW th respectively. The scheme of the system is shown. The fast core consists of 48 hexagonal fuel bundles each containing 169 fuel pins of 8.2 mm diameter arranged in 11.4 mm triangular array pitch. The average thermal power per fuel pin is about 13.46 kw. However, due to neutron flux peaking effect, the maximum fuel pin power can be up to 2.5 times this average power. The thermal reactor consists of heavy water as moderator and coolant similar to a typical CANDU type Indian PHWR except for fuel composition. Though the gap between fast and thermal zones essentially provides one way coupling of neutron flux, a thermal

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Brayton rotating units for space reactor power systems

Energy Technology Data Exchange (ETDEWEB)

Gallo, Bruno M.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept., The Univ. of New Mexico, Albuquerque, NM 87131 (United States)

2009-09-15

Designs and analyses models of centrifugal-flow compressor and radial-inflow turbine of 40.8kW{sub e} Brayton Rotating Units (BRUs) are developed for 15 and 40 g/mole He-Xe working fluids. Also presented are the performance results of a space power system with segmented, gas cooled fission reactor heat source and three Closed Brayton Cycle loops, each with a separate BRU. The calculated performance parameters of the BRUs and the reactor power system are for shaft rotational speed of 30-55 krpm, reactor thermal power of 120-471kW{sub th}, and turbine inlet temperature of 900-1149 K. With 40 g/mole He-Xe, a power system peak thermal efficiency of 26% is achieved at rotation speed of 45 krpm, compressor and turbine inlet temperatures of 400 and 1149 K and 0.93 MPa at exit of the compressor. The corresponding system electric power is 122.4kW{sub e}, working fluid flow rate is 1.85 kg/s and the pressure ratio and polytropic efficiency are 1.5% and 86.3% for the compressor and 1.42% and 94.1% for the turbine. For the same nominal electrical power of 122.4kW{sub e}, decreasing the molecular weight of the working fluid (15 g/mole) decreases its flow rate to 1.03 kg/s and increases the system pressure to 1.2 MPa. (author)

Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

International Nuclear Information System (INIS)

Koenig, D.R.; Gido, R.G.; Brandon, D.I.

1985-01-01

The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

Development of fluid system design technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, D. J.; Chang, M. H.; Kang, D. J. and others

1999-03-01

This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

International Nuclear Information System (INIS)

Inaba, Yoshitomo

2008-03-01

Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

1977-01-01

A new type of superconducting magnet system for large fusion reactors is described in this report. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil

Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

1978-01-01

A new type of superconducting magnet system for large fusion reactors is described. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil. (author)

A severe accident analysis for the system-integrated modular advanced reactor

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

How Accurately can we Calculate Thermal Systems?

International Nuclear Information System (INIS)

Cullen, D; Blomquist, R N; Dean, C; Heinrichs, D; Kalugin, M A; Lee, M; Lee, Y; MacFarlan, R; Nagaya, Y; Trkov, A

2004-01-01

I would like to determine how accurately a variety of neutron transport code packages (code and cross section libraries) can calculate simple integral parameters, such as K eff , for systems that are sensitive to thermal neutron scattering. Since we will only consider theoretical systems, we cannot really determine absolute accuracy compared to any real system. Therefore rather than accuracy, it would be more precise to say that I would like to determine the spread in answers that we obtain from a variety of code packages. This spread should serve as an excellent indicator of how accurately we can really model and calculate such systems today. Hopefully, eventually this will lead to improvements in both our codes and the thermal scattering models that they use in the future. In order to accomplish this I propose a number of extremely simple systems that involve thermal neutron scattering that can be easily modeled and calculated by a variety of neutron transport codes. These are theoretical systems designed to emphasize the effects of thermal scattering, since that is what we are interested in studying. I have attempted to keep these systems very simple, and yet at the same time they include most, if not all, of the important thermal scattering effects encountered in a large, water-moderated, uranium fueled thermal system, i.e., our typical thermal reactors

Thermal shielding device in LMFBR type reactors

International Nuclear Information System (INIS)

Nakamura, Hiroshi.

1985-01-01

Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

Modernization of turbine control system and reactor control system in Almaraz 1 and 2

International Nuclear Information System (INIS)

Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

2005-01-01

The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

Method and apparatus for a combination moving bed thermal treatment reactor and moving bed filter

Energy Technology Data Exchange (ETDEWEB)

Badger, Phillip C.; Dunn, Jr., Kenneth J.

2015-09-01

A moving bed gasification/thermal treatment reactor includes a geometry in which moving bed reactor particles serve as both a moving bed filter and a heat carrier to provide thermal energy for thermal treatment reactions, such that the moving bed filter and the heat carrier are one and the same to remove solid particulates or droplets generated by thermal treatment processes or injected into the moving bed filter from other sources.

Revision of construction plan for advanced thermal demonstration reactor

International Nuclear Information System (INIS)

1996-01-01

The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)

A Novel Thermal-Mechanical Detection System for Reactor Pressure Vessel Bottom Failure Monitoring in Severe Accidents

International Nuclear Information System (INIS)

Bi, Daowei; Bu, Jiangtao; Xu, Dongling

2013-06-01

Following the Fukushima Daiichi nuclear accident in Japan, there is an increased need of enhanced capabilities for severe accident management (SAM) program. Among others, a reliable method for detecting reactor pressure vessel (RPV) bottom failure has been evaluated as imperative by many utility owners. Though radiation and/or temperature measurement are potential solutions by tradition, there are some limitations for them to function desirably in such severe accident as that in Japan. To provide reliable information for assessment of accident progress in SAM program, in this paper we propose a novel thermal-mechanical detection system (TMDS) for RPV bottom failure monitoring in severe accidents. The main components of TMDS include thermally sensitive element, metallic cables, tension controlled switch and main control room annunciation device. With TMDS installed, there shall be a reliable means of keeping SAM decision-makers informed whether the RPV bottom has indeed failed. Such assurance definitely guarantees enhancement of severe accident management performance and significantly improve nuclear safety and thus protect the society and people. (authors)

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

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

International Nuclear Information System (INIS)

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

1982-01-01

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

A cermet fuel reactor for nuclear thermal propulsion

Science.gov (United States)

Kruger, Gordon

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

Thermal analysis of CANDU6 moderator system for loss of cooling

International Nuclear Information System (INIS)

Xu Zhen

2012-01-01

The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, the thermal analysis of moderator system for loss of cooling was compared with the experiment data and the system failure caused by the temperature raising was evaluated in this paper. (author)

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

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

1998-01-01

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

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Backfitting in Rossendorf research reactor control and instrumentation system

International Nuclear Information System (INIS)

Klebau, J.; Seidler, S.

1985-01-01

The paper generally describes a decentralized Hierarchical Information System (HIS) which has been developed for backfitting in Rossendorf Research Reactor (RFR) control and instrumentation system. The RFR was put into operation in 1957 and reconstructed from 2 MW up to a thermal power of 10 MW at the end of the sixties. Backfitting is planned by use of an advanced computerized control system for the next years. Main tasks of HIS are: Processmonitoring, online-disturbance analysis, technical diagnosis, direct digital control and use of a special industrial robot for discharging of irradiated materials out of the reactor. Experiences obtained by HIS during a testperiod will be presented. (author)

Thermal and mechanical behaviour of oxygen carrier materials for chemical looping combustion in a packed bed reactor

International Nuclear Information System (INIS)

Jacobs, M.; Van Noyen, J.; Larring, Y.; Mccann, M.; Pishahang, M.; Amini, S.; Ortiz, M.; Galluci, F.; Sint-Annaland, M.V.; Tournigant, D.; Louradour, E.; Snijkers, F.

2015-01-01

Highlights: • Ilmenite-based oxygen carriers were developed for packed-bed chemical looping. • Addition of Mn_2O_3 increased mechanical strength and microstructure of the carriers. • Oxygen carriers were able to withstand creep and thermal cycling up to 1200 °C. • Ilmenite-based granules are a promising shape for packed-bed reactor conditions. - Abstract: Chemical looping combustion (CLC) is a promising carbon capture technology where cyclic reduction and oxidation of a metallic oxide, which acts as a solid oxygen carrier, takes place. With this system, direct contact between air and fuel can be avoided, and so, a concentrated CO_2 stream is generated after condensation of the water in the exit gas stream. An interesting reactor system for CLC is a packed bed reactor as it can have a higher efficiency compared to a fluidized bed concept, but it requires other types of oxygen carrier particles. The particles must be larger to avoid a large pressure drop in the reactor and they must be mechanically strong to withstand the severe reactor conditions. Therefore, oxygen carriers in the shape of granules and based on the mineral ilmenite were subjected to thermal cycling and creep tests. The mechanical strength of the granules before and after testing was investigated by crush tests. In addition, the microstructure of these oxygen particles was studied to understand the relationship between the physical properties and the mechanical performance. It was found that the granules are a promising shape for a packed bed reactor as no severe degradation in strength was noticed upon thermal cycling and creep testing. Especially, the addition of Mn_2O_3 to the ilmenite, which leads to the formation of an iron–manganese oxide, seems to results in stronger granules than the other ilmenite-based granules.

Validation of reactor core protection system

International Nuclear Information System (INIS)

Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

2008-01-01

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

Development of intellectual reactor design system IRDS

International Nuclear Information System (INIS)

Kugo, T.; Tsuchihashi, K.; Nakagawa, M.; Mori, T.

1993-01-01

An intellectual reactor design system IRDS has been developed to support feasibility study and conceptual design of new type reactors in the fields of reactor core design including neutronics, thermal-hydraulics and fuel design. IRDS is an integrated software system in which a variety of computer codes in the different fields are installed. An integration of simulation modules are performed by the information transfer between modules through design model in which the design information of the current design work is stored. An object oriented architecture is realized in frame representation of core configuration in a design data base. The knowledge relating to design tasks to be performed are encapsulated, to support the conceptual design work. The system is constructed on an engineering workstation, and supports efficiently design work through man-machine interface adopting the advanced information processing technologies. Optimization methods for design parameters with use of the artificial intelligence technique are now under study, to reduce the parametric study work. A function to search design window in which design is feasible is realized in the fuel pin design. (orig.)

Analysis of Russian transition scenarios to innovative nuclear energy system based on thermal and fast reactors with closed nuclear fuel cycle using INPRO methodology

International Nuclear Information System (INIS)

Kagramanyan, V.S.; Poplavskaya, E.V.; Korobeynikov, V.V.; Kalashnikov, A.G.; Moseev, A.L.; Korobitsyn, V.E.; Andreeva-Andrievskaya, L.N.

2011-01-01

This paper presents the results of the analysis of modeling of Russian nuclear energy (NE) scenarios on the basis of thermal and fast reactors with closed nuclear fuel cycle (NFC). Modeling has been carried out with use of CYCLE code (SSC RF IPPE's tool) designed for analysis of Nuclear Energy System (NES) with closed NFC taking into account plutonium and minor actinides (MA) isotopic composition change during multi-recycling of fuel in fast reactors. When considering fast reactor introduction scenarios, one of important questions is to define optimal time for their introduction and related NFC's facilities. Analysis of the results obtained has been fulfilled using the key INPRO indicators for sustainable energy development. It was shown that a delay in fast reactor introduction led to serious ecological, social and finally economic risks for providing energy security and sustainable development of Russia in long-term prospects and loss of knowledge and experience in mastering innovative technologies of fast reactors and related nuclear fuel cycle. (author)

Natural circulating passive cooling system for nuclear reactor containment structure

Science.gov (United States)

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

International Nuclear Information System (INIS)

Kim, Hak Jae; Park, Goon Cherl

1996-01-01

Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

Reactor pressure vessel thermal annealing

International Nuclear Information System (INIS)

Lee, A.D.

1997-01-01

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

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

Preliminary design concepts for the advanced neutron source reactor systems

International Nuclear Information System (INIS)

Peretz, F.J.

1988-01-01

This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

International Nuclear Information System (INIS)

Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

1994-05-01

The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

Physics of Plutonium Recycling in Thermal Reactors

International Nuclear Information System (INIS)

Kinchin, G.H.

1967-01-01

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

Physics of Plutonium Recycling in Thermal Reactors

Energy Technology Data Exchange (ETDEWEB)

Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1967-09-15

A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

New reactor technology: safety improvements in nuclear power systems.

Science.gov (United States)

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Computational analysis of battery optimized reactor integral system

International Nuclear Information System (INIS)

Hwang, J. S.; Son, H. M.; Jeong, W. S.; Kim, T. W.; Suh, K. Y.

2007-01-01

Battery Optimized Reactor Integral System (BORIS) is being developed as a multi-purpose fast spectrum reactor cooled by lead (Pb). BORIS is an integral optimized reactor with an ultra-long life core. BORIS aims to satisfy various energy demands maintaining inherent safety with the primary coolant Pb, and improving economics. BORIS is being designed to generate 23 MW t h with 10 MW e for at least twenty consecutive years without refueling and to meet the Generation IV Nuclear Energy System goals of sustainability, safety, reliability, and economics. BORIS is conceptualized to be used as the main power and heat source for remote areas and barren lands, and also considered to be deployed for desalinisation purpose. BORIS, based on modular components to be viable for rapid construction and easy maintenance, adopts an integrated heat exchanger system operated by natural circulation of Pb without pumps to realize a small sized reactor. The BORIS primary system is designed through an optimization study. Thermal hydraulic characteristics during a reactor steady state with heat source and sink by core and heat exchanger, respectively, have been carried out by utilizing a computational fluid dynamics code and hand calculations based on first principles. This paper analyzes a transient condition of the BORIS primary system. The Pb coolant was selected for its lower chemical activity with air or water than sodium (Na) and good thermal characteristics. The reactor transient conditions such as core blockage, heat exchanger failure, and loss of heat sink, were selected for this study. Blockage in the core or its inlet structure causes localized flow starvation in one or several fuel assemblies. The coolant loop blockages cause a more or less uniform flow reduction across the core, which may trigger coolant temperature transient. General conservation equations were applied to model the primary system transients. Numerical approaches were adopted to discretized the governing

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Thermal-hydraulic methods in fast reactor safety

International Nuclear Information System (INIS)

Weber, D.P.; Briggs, L.L.

1985-01-01

Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

2002-08-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

Phenomena occurring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1989-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Startup thaw concept for the SP-100 space reactor power system

Science.gov (United States)

Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

1990-01-01

A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

Thermal-Hydraulic Analysis of a Supercritical Water Reactor (SCWR) Core

International Nuclear Information System (INIS)

Kucukboyaci, V.N.; Oriani, L.

2004-01-01

The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor

Thorium utilization: conversion ratio and fuel needs in thermal reactors

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1975-01-01

As a preparatory study for thorium utilization in thermal reactors a study has been made of the fuel comsumption in existing reactor types. A quantitative description is given of the influence of enrichment, burnup, amount of structural material, choise of coolant and control requirements on the convertion ratio. The enrichment is an important factor and a low fuel comsumption can be achieved by increasing the enrichment

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1980-01-01

This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes. Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved. (orig.)

Coupled CFD - system-code simulation of a gas cooled reactor

International Nuclear Information System (INIS)

Yan, Yizhou; Rizwan-uddin

2011-01-01

A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

2013-01-01

In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

Nuclear reactor, fuel assembly and neutron measuring system

International Nuclear Information System (INIS)

Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

1998-01-01

The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)

2000-10-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu

2000-01-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

International Nuclear Information System (INIS)

Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

1989-01-01

The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

Nuclear reactor insulation and preheat system

International Nuclear Information System (INIS)

Wampole, N.C.

1978-01-01

An insulation and preheat system is disclosed for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the ocmpartment. An external surface of the compartment of enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

1981-01-01

The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

International Nuclear Information System (INIS)

1983-06-01

This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

Energy Technology Data Exchange (ETDEWEB)

Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

2016-09-15

This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

International Nuclear Information System (INIS)

Cuevas V, G.; Banfield, J.; Avila N, A.

2016-09-01

This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

Integrated system of nuclear reactor and heat exchanger

International Nuclear Information System (INIS)

McDonald, B.N.; Schluderberg, D.C.

1977-01-01

The invention concerns PWRs in which the heat exchanger is associated with a pressure vessel containing the core and from which it can be selectively detached. This structural configuration applies to electric power generating uses based on land or on board ships. An existing reactor of this kind is fitted with a heat exchanger in which the tubes are 'U' shaped. This particular design of heat exchangers requires that the ends of the curved tubes be solidly maintained in a tube plate of great thickness, hence difficult to handle and to fabricate and requiring unconventional fine control systems for the control rods and awkward coolant pump arrangements. These complications limit the thermal power of the system to level below 100 megawatts. On the contrary, the object of this invention is to provide a one-piece PWR reactor capable of reaching power levels of 1500 thermal megawatts at least. For this, a pressure vessel is provided in the cylindrical assembly with not only a transversal separation on a plane located between the reactor and the heat exchanger but also a cover selectively detachable which supports the fine control gear of the control rods. Removing the cover exposes a part of the heat exchanger for easy inspection and maintenance. Further, the heat exchanger can be removed totally from the pressure vessel containing the core by detaching the cylindrical part, which composes the heat exchanger section, from the part that holds the reactor core on a level with the transversal separation [fr

Study on thermal neutron spectra in reactor moderators by time-of-flight method

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-12-01

Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

International Nuclear Information System (INIS)

Toti, A.; Vierendeels, J.; Belloni, F.

2017-01-01

Highlights: • A system thermal-hydraulic/CFD coupling methodology is proposed for high-fidelity transient flow analyses. • The method is based on domain decomposition and implicit numerical scheme. • A novel interface Quasi-Newton algorithm is implemented to improve stability and convergence rate. • Preliminary validation analyses on the TALL-3D experiment. - Abstract: The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK• CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.

Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

International Nuclear Information System (INIS)

Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

1980-09-01

The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

Improved measurements of thermal power and control rods using multiple detectors at the TRIGA Mark II reactor in Ljubljana

International Nuclear Information System (INIS)

Zerovnik Gasper; Snoj Luka; Trkov Andrej; Barbot Loic; Fourmentel Damien; Villard Jean-Francois

2013-06-01

The aim of the current bilateral project between CEA Cadarache and JSI is to improve the accuracy of the online thermal power monitoring at the JSI TRIGA reactor. Simultaneously, a new wide range multichannel acquisition system for fission chambers, recently developed by CEA, is tested. In the paper, calculational and experimental power calibration methods are described. The focus is on use of multiple detectors in combination with pre-calculated and pre-measured control rod- position-dependent correction factors to improve the reactor power reading. The system will be implemented and tested at the JSI TRIGA reactor in 2014. (authors)

Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Wosko, Paul; Sundram, S. K.

2012-10-16

New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

Survey of thermal insulation systems

International Nuclear Information System (INIS)

Kinoshita, Izumi

1983-01-01

Better thermal insulations have been developed to meet the growing demands of industry, and studies on thermal insulation at both high temperature and low temperature have been widely performed. The purpose of this survey is to summarize data on the performances and characteristics of thermal insulation materials and thermal insulation structures (for instance, gas cooled reactors, space vehicles and LNG storage tanks), and to discuss ravious problems regarding the design of thermal insulation structures of pool-type LMFBRs. (author)

Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

Energy Technology Data Exchange (ETDEWEB)

Bernstein, A; Bowden, N; Misner, A; Palmer, T

2007-06-27

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7 days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors

International Nuclear Information System (INIS)

Williams, M.M.R.

2007-01-01

Description: Prof. M.M..R Williams has now released three of his legacy books for free distribution: 1 - M.M.R. Williams: The Slowing Down and Thermalization of Neutrons, North-Holland Publishing Company - Amsterdam, 582 pages, 1966. Content: Part I - The Thermal Energy Region: 1. Introduction and Historical Review, 2. The Scattering Kernel, 3. Neutron Thermalization in an Infinite Homogeneous Medium, 4. Neutron Thermalization in Finite Media, 5. The Spatial Dependence of the Energy Spectrum, 6. Reactor Cell Calculations, 7. Synthetic Scattering Kernels. Part II - The Slowing Down Region: 8. Scattering Kernels in the Slowing Down Region, 9. Neutron Slowing Down in an Infinite Homogeneous Medium, 10.Neutron Slowing Down and Diffusion. 2 - M.M.R. Williams: Mathematical Methods in Particle Transport Theory, Butterworths, London, 430 pages, 1971. Content: 1 The General Problem of Particle Transport, 2 The Boltzmann Equation for Gas Atoms and Neutrons, 3 Boundary Conditions, 4 Scattering Kernels, 5 Some Basic Problems in Neutron Transport and Rarefied Gas Dynamics, 6 The Integral Form of the Transport Equation in Plane, Spherical and Cylindrical Geometries, 7 Exact Solutions of Model Problems, 8 Eigenvalue Problems in Transport Theory, 9 Collision Probability Methods, 10 Variational Methods, 11 Polynomial Approximations. 3 - M.M.R. Williams: Random Processes in Nuclear Reactors, Pergamon Press Oxford New York Toronto Sydney, 243 pages, 1974. Content: 1. Historical Survey and General Discussion, 2. Introductory Mathematical Treatment, 3. Applications of the General Theory, 4. Practical Applications of the Probability Distribution, 5. The Langevin Technique, 6. Point Model Power Reactor Noise, 7. The Spatial Variation of Reactor Noise, 8. Random Phenomena in Heterogeneous Reactor Systems, 9. Associated Fluctuation Problems, Appendix: Noise Equivalent Sources. Note to the user: Prof. M.M.R Williams owns the copyright of these books and he authorises the OECD/NEA Data Bank

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

International Nuclear Information System (INIS)

Corradin, Michael; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

2014-01-01

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

Energy Technology Data Exchange (ETDEWEB)

Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

2014-10-15

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

Deposition reactors for solar grade silicon: A comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

Science.gov (United States)

Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.

2015-12-01

Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.

Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors

International Nuclear Information System (INIS)

Scervini, M.; Rae, C.; Lindley, B.

2013-06-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath

Application of Cherenkov light observation to reactor measurements (2). Design and trial fabrication of Cherenkov light estimation system

International Nuclear Information System (INIS)

Yamamoto, Keiichi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Hayashi, Takayasu; Kosuge, Fumiaki; Sano, Tadafumi

2015-11-01

Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. This report summarized the results of design and trial fabrication of the Cherenkov light estimation system from thermal power evaluation from Cherenkov light image emitted from the fuel elements. The developed Cherenkov light estimation system was verified with the Cherenkov light image emitted from the fuels in the core of Kyoto University Research Reactor (KUR). From the results, the thermal power of the fuel elements evaluated from the brightness of the Cherenkov light observed by a CCD camera was almost the same as that of thermal power calculated from SRAC code. On the other hand, the evaluation values of some fuel elements were different from the calculation values. This, it is necessary to improve the observation method of Cherenkov light in the reactor and the evaluation method of the brightness of Cherenkov light. (author)

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

Directory of Open Access Journals (Sweden)

Fic Adam

2015-03-01

Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.