Advanced gas cooled nuclear reactor materials evaluation and development program

International Nuclear Information System (INIS)

1977-01-01

Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

Status of experimental data for the VHTR core design

Energy Technology Data Exchange (ETDEWEB)

Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

2004-05-01

The VHTR (Very High Temperature Reactor) is being emerged as a next generation nuclear reactor to demonstrate emission-free nuclear-assisted electricity and hydrogen production. The VHTR could be either a prismatic or pebble type helium cooled, graphite moderated reactor. The final decision will be made after the completion of the pre-conceptual design for each type. For the pre-conceptual design for both types, computational tools are being developed. Experimental data are required to validate the tools to be developed. Many experiments on the HTGR (High Temperature Gas-cooled Reactor) cores have been performed to confirm the design data and to validate the design tools. The applicability and availability of the existing experimental data have been investigated for the VHTR core design in this report.

Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

Energy Technology Data Exchange (ETDEWEB)

Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

International Nuclear Information System (INIS)

Limaiem, I.

2006-12-01

Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

Guidelines for the structural design of experimental multi-purpose VHTR at the elevated temperature services

International Nuclear Information System (INIS)

Nomura, Sueo; Uga, Takeo; Miyamoto, Yoshiaki; Muto, Yasushi; Ikushima, Takeshi

1976-02-01

The guidelines are presented for structural design of the experimental multi-purpose VHTR(Very High Temperature Reactor) at the elevated temperature services. Covered are features of the VHTR structural design, specifications, safety design, seismic design, failure modes to be considered, stress criteria for various load combinations and the mechanical properties of the materials. The guidelines were prepared by referring to safety criteria of high-temperature gas cooled reactors, ASME Boiler and Pressure Vessel code, Section III, case 1592 and the domestic seismic design guide of nuclear power facilities. (auth.)

Current Status of the Elevated Temperature Structure Design Codes for VHTR

International Nuclear Information System (INIS)

Kim, Jong-Bum; Kim, Seok-Hoon; Park, Keun-Bae; Lee, Won-Jae

2006-01-01

An elevated temperature structure design and analysis is one of the key issues in the VHTR (Very High Temperature Reactor) project to achieve an economic production of hydrogen which will be an essential energy source for the near future. Since the operating temperature of a VHTR is above 850 .deg. C, the existing code and standards are insufficient for a high temperature structure design. Thus the issues concerning a material selection and behaviors are being studied for the main structural components of a VHTR in leading countries such as US, France, UK, and Japan. In this study, the current status of the ASME code, French RCC-MR, UK R5, and Japanese code were investigated and the necessary R and D items were discussed

Studies on the core-support carbon material for VHTR, (1)

International Nuclear Information System (INIS)

Matsuo, Hideto; Saito, Tamotsu; Fukuda, Yasumasa; Sasaki, Yasuichi; Hasegawa, Takashi.

1979-11-01

To obtain information of core-support carbon material for VHTR, thermal conductivity and electrical resistivity of three domestic carbon blocks were measured. Results indicated the need for development of carbon material with lower thermal conductivity for VHTR. These two were also measured of the samples heat-treated between 1000 0 C and 3040 0 C for one hour. Thermal conductivity increased with heat-treatment above 1200 0 C and resistivity stayed constant between 1500 0 C and 2000 0 C. The results should be useful in choosing the final heat-treatment temperature in carbon material production. The changes of Lorentz number with heat treatment were classified into three heat-treatment temperature regions of below 1500 0 C, 1500 0 C - 2500 0 C, and above 2500 0 C; the results are interpreted with a graphitization model. (author)

Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

International Nuclear Information System (INIS)

Wang, Lumin; Was, Gary

2010-01-01

Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

Experimental Measurement of Flow Phenomena in a VHTR Lower Plenum Model

Energy Technology Data Exchange (ETDEWEB)

Hugh M. McIlroy Jr.; Keith G. Condie; Glenn E. McCreery; Donald M. McEligot; Robert J. Pink

2006-06-01

The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligible buoyancy and constant fluid properties.

VHTR engineering design study: intermediate heat exchanger program. Final report

International Nuclear Information System (INIS)

1976-11-01

The work reported is the result of a follow-on program to earlier Very High Temperature Reactor (VHTR) studies. The primary use of the VHTR is to provide heat for various industrial processes, such as hydrocarbon reforming and coal gasification. For many processes the use of an intermediate heat transfer barrier between the reactor coolant and the process is desirable; for some processes it is mandatory. Various intermediate heat exchanger (IHX) concepts for the VHTR were investigated with respect to safety, cost, and engineering design considerations. The reference processes chosen were steam-hydrocarbon reforming, with emphasis on the chemical heat pipe, and steam gasification of coal. The study investigates the critically important area of heat transfer between the reactor coolant, helium, and the various chemical processes

Development on experimental VHTR instrumentation

International Nuclear Information System (INIS)

Wakayama, N.; Ara, K.; Terada, H.; Yamagishi, H.; Tomoda, T.

1982-06-01

This paper describes developmental works on the instrumentation of the Experimental VHTR. In the area of the nuclear instrumentation for the reactor control, high temperature fission counter-chambers have been developed. These withstood the accelerated irradiation life tests at 600 deg. C, the long term in-reactor operating test at 600 deg. C and the 800 deg. C-operating tests for several hundred hours in a simulated accident condition. Platinum-Molybdenum alloy thermocouples have been studied as a neutron-irradiation-resistant high-temperature thermocouple for the in-core temperature distribution monitoring of the VHTR in the temperature range between 1000 deg. C and 1350 deg. C. The instability problems of the Pt-5% Mo/Pt-0.1% Mo thermocouple seem to be overcome by introducing a double sheath structure and adopting a better material to the inner sheath. A local failure and abnormality monitoring method for the HTR fuel is also studied using a gas-sweeping irradiation rig for the CPF compacts. This study aims mainly at the development of a method to compensate for the dependency of the FP-release rate on the fuel temperature, the neutron flux density, the burn-up and others, in order to increase the detection sensitivity of fuel failures. (author)

Spectral emissivity measurements of candidate materials for very high temperature reactors

Energy Technology Data Exchange (ETDEWEB)

Cao, G.; Weber, S.J.; Martin, S.O.; Anderson, M.H. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Sridharan, K., E-mail: kumars@cae.wisc.edu [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States)

2012-10-15

Heat dissipation by radiation is an important consideration in VHTR components, particularly the reactor pressure vessel (RPV), because of the fourth power temperature dependence of radiated heat. Since emissivity is the material property that dictates the ability to radiate heat, measurements of emissivities of materials that are being specifically considered for the construction of VHTR become important. Emissivity is a surface phenomenon and therefore compositional, structural, and topographical changes that occur at the surfaces of these materials as a result of their interactions with the environment at high temperatures will alter their emissivities. With this background, an experimental system for the measurement of spectral emissivity has been designed and constructed. The system has been calibrated in conformance with U.S. DoE quality assurance standards using inert ceramic materials, boron nitride, silicon carbide, and aluminum oxide. The results of high temperature emissivity measurements of potential VHTR materials such as ferritic steels SA 508, T22, T91 and austenitic alloys IN 800H, Haynes 230, IN 617, and 316 stainless steel have been presented.

VHTR Construction Ripple Effect using Inter-Industry Analysis

International Nuclear Information System (INIS)

Lee, T. H.; Lee, K. Y.; Shin, Y. J.

2015-01-01

As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the exact cost due to insufficient reference data and experience. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt x 4 module construction and operation ripple effect based on NOAK. This paper presents a new method for the ripple effect and preliminary ripple effect consequence. We proposed a ripple effect analysis method using a time series and inter-industry table. As a result, we can predict that a 600MWth x 4 module VHTR reactor construction will bring about a 43771 employment effect, 24160 billion KRW production effect, and 4472 billion added value effect for 22 years. It is necessary to use the sub-account values of an inter-industry table to obtain a more precise effect result. However, the methodology can be applied with minor modification to another reactor type.

VHTR Construction Ripple Effect using Inter-Industry Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, T. H.; Lee, K. Y.; Shin, Y. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the exact cost due to insufficient reference data and experience. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt x 4 module construction and operation ripple effect based on NOAK. This paper presents a new method for the ripple effect and preliminary ripple effect consequence. We proposed a ripple effect analysis method using a time series and inter-industry table. As a result, we can predict that a 600MWth x 4 module VHTR reactor construction will bring about a 43771 employment effect, 24160 billion KRW production effect, and 4472 billion added value effect for 22 years. It is necessary to use the sub-account values of an inter-industry table to obtain a more precise effect result. However, the methodology can be applied with minor modification to another reactor type.

Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

Energy Technology Data Exchange (ETDEWEB)

Fok, Alex

2013-10-30

The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

Strategy of VHTR Realization

Energy Technology Data Exchange (ETDEWEB)

Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day.

Strategy of VHTR Realization

International Nuclear Information System (INIS)

Chang, Jonghwa

2015-01-01

High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day

A Review on the VHTR PIRT Development Status of Both Regulatory Authority and Licensee

International Nuclear Information System (INIS)

Hwang, Su Hyun; Jeon, Seong Su; Hong, Soon Joon; Lee, Byung Chul; Huh, Chang Wook; Jin, Chang Yong; Kim, Kyun Tae

2011-01-01

The VHTR (Very High Temperature Reactor) is defined as a helium-cooled, graphite moderated reactor with a core outlet temperature in excess of 900 .deg. C and a long-term goal of achieving an outlet temperature of 1000 .deg. C. The VHTR is suited for a broad range of applications, including the production of hydrogen and electricity. The PIRT (Phenomena Identification and Ranking Table) provides a structured means of identifying and analyzing a wide variety of off-normal sequences that potentially challenge the viability of complex technological systems. As applied to VHTR, the PIRT is used to identify a spectrum of safety-related sequences or phenomena that could affect those systems, and to rank order those sequences on the basis of their frequencies, their potential consequences, and state of knowledge related to associate phenomena. It is to be used as an early screening tool to identify, categorize, and characterize phenomena and issues that are potentially important to risk and safety of VHTR. Since a specific design has not yet been selected for the choice of the US VHTR (NGNP), it was decided early on to focus on a generic plant and reactor design with broadly typical features. Both a generic Pebble Bed Reactor (PBR) design and a generic Prismatic Modular Reactor (PMR) design were selected as the reference plant for KAERI and ANL PIRT. The generic PBR design selected is a version of the 400 MWt South African PBMR design. The generic PMR design selected is a version of the 600 MWt GT-MHR. The reference plant of NRC PIRT is assumed to be a modular high temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GT-MHR) version (a prismatic-core modular reactor- PMR) or a pebble bed modular reactor (PBMR) version (a pebble bed reactor-PBR) design, with either a director indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. The difference of VHTR PIRT

Assessment of very high temperature reactors in process applications

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

1976-01-01

In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

International Nuclear Information System (INIS)

Jo, Chang Keun; Noh, Jae Man

2007-01-01

This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving VHTR Efficiency and Testing Material Compatibility - Final Report

International Nuclear Information System (INIS)

Chang H. Oh

2006-01-01

Generation IV reactors will need to be intrinsically safe, having a proliferation-resistant fuel cycle and several advantages relative to existing light water reactor (LWR). They, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30 percent reduction in power cost for state-of-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to the Very-High-Temperature Gas-Cooled Reactor (VHTR), (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase plant net efficiency

ANTARES: The HTR/VHTR project at Framatome ANP

International Nuclear Information System (INIS)

Gauthier, Jean-Claude; Brinkmann, Gerd; Copsey, Bernie; Lecomte, Michel

2006-01-01

Framatome ANP is developing a very high temperature reactor (VHTR), relying on its previous experience with high temperature reactor concepts, from its participation in the MODUL and the GT-MHR designs. While being a major actor in the nuclear reactor business with proven light water technology, AREVA wishes to be ready to meet the new challenges calling for small grid requirements, high temperature process heat and cogeneration. The Framatome ANP VHTR design for electricity production is based on an indirect cycle coupled to an 'off-the-shelf' combined cycle gas turbine. Although direct cycle HTRs are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding power generation system (PGS) developments and keeping the PGS contamination free. Moreover, the nuclear heat source of the indirect cycle could also be used to meet the heat supplies from a standard design for multiple applications

Evaluation of the DRAGON code for VHTR design analysis.

Energy Technology Data Exchange (ETDEWEB)

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

Evaluation of the DRAGON code for VHTR design analysis

International Nuclear Information System (INIS)

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-01

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

A reliability assessment methodology for the VHTR passive safety system

International Nuclear Information System (INIS)

Lee, Hyungsuk; Jae, Moosung

2014-01-01

The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

Preliminary Estimation of Local Bypass Flow Gap Sizes for a Prismatic VHTR Core

International Nuclear Information System (INIS)

Kim, Min Hwan; Jo, Chang Keun; Lee, Won Jae

2009-01-01

The Very High Temperature Reactor (VHTR) has been selected for the Nuclear Hydrogen Development and Demonstration (NHDD) project. In the VHTR design, core bypass flow has been one of key issues for core thermal margins and target temperature of the core outlet. The core bypass flow in the prismatic VHTR varies with the core life due to the irradiation shrinkage/ swelling and thermal expansion of the graphite blocks, which could be a significant proportion of the total core flow. Thus, accurate prediction of the bypass flow is of major importance in assuring the core thermal margin. To predict the bypass flow, first of all, local gap sizes between graphite blocks in the core should be determined. The objectives of this work are to develop a methodology for determining the gap sizes and to perform a preliminary evaluation for a reference reactor

Safety assessment of VHTR hydrogen production system against fire, explosion and acute toxicity

International Nuclear Information System (INIS)

Murakami, Tomoyuki; Nishihara, Tetsuo; Kunitomi, Kazuhiko

2008-01-01

The Japan Atomic Energy Agency has been developing a nuclear hydrogen production system by using heat from the Very High Temperature Reactor (VHTR). This system will handle a large amount of combustible gas and toxic gas. The risk from fire, explosion and acute toxic exposure caused by an accident involving chemical material release in a hydrogen production system is assessed. It is important to ensure the safety of the nuclear plant, and the risks for public health should be sufficiently small. This report provides the basic policy for the safety evaluation in cases of accident involving fire, explosion and toxic material release in a hydrogen production system. Preliminary safety analysis of a commercial-sized VHTR hydrogen production system, GTHTR300C, is performed. This analysis provides us with useful information on the separation distance between a nuclear plant and a hydrogen production system and a prospect that an accident in a hydrogen production system does not significantly increase the risks of the public. (author)

Neutronic performance of high molecular weight coolants for a prismatic VHTR

International Nuclear Information System (INIS)

Schriener, T. M.; El-Genk, M. S.

2008-01-01

A neutronic model is developed of a prismatic Very High Temperature Reactor (VHTR) to investigate the effects on the excess reactivity and operation cycle length of replacing helium with binary gas mixtures of He-Ne, He-N 2 , or He-Xe as reactor coolants and working fluids in the direct Closed Brayton Cycle (CBC) for energy conversion. Also investigated is the neutron activation of these binary gas mixtures in the VHTR. The motivation for using the heavy binary mixtures is the smaller size and the fewer number of stages of the CBC turbo-machinery. The present analysis uses the Monte Carlo code MCNPX 2.6D at typical operating conditions (500-1000 degrees and 7.12 MPa) in the VHTR. He-Ne (15 g/mol) is the best neutronically, but not thermal-hydraulically, followed by He-N 2 . Although He-Ne has ∼13.6% lower heat transfer coefficient than helium, it insignificantly affects the initial excess reactivity and the operation life cycle and experiences no neutrons activation. On the other hand, He-N 2 has 4.4% higher heat transfer coefficient than helium and experiences insignificant neutron activation in the reactor, but decreases the initial excess reactivity by ∼5.2% and the operation cycle length by 6.7%. He-Xe (15 g/mol) has 8% higher heat transfer coefficient than helium, but decreases the initial excess reactivity by 18.2% and the operational cycle length by 17%. In addition, neutron activation of xenon produces a significant source term, requiring shielding of the CBC loop and could contaminate the turbo-machinery with long-lived radioactive cesium. Thus, He-Xe is not recommended as a reactor coolant, but could be used as working fluid in a CBC loop that is indirectly coupled to helium cooled VHTR. (authors)

Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

Science.gov (United States)

Valentin Rodriguez, Francisco Ivan

High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

Development of flow network analysis code for block type VHTR core by linear theory method

International Nuclear Information System (INIS)

Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

2012-01-01

VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

State of the Art Report for a Bearing for VHTR Helium Circulator

International Nuclear Information System (INIS)

Lee, Jae Seon; Song, Kee Nam; Kim, Yong Wan; Lee, Won Jae

2008-10-01

A helium circulator in a VHTR(Very High Temperature gas-cooled Reactor) plays a core role which translates thermal energy at high temperature from a nuclear core to a steam generator. Helium as a operating coolant circulates a primary circuit in high temperature and high pressure state, and controls thermal output of a nuclear core by controlling flow rate. A helium circulator is the only rotating machinery in a VHTR, and its reliability should be guaranteed for reliable operation of a reactor and stable production of hydrogen. Generally a main helium circulator is installed on the top of a steam generator vessel, and helium is circulated only by a main helium circulator in a normal operation state. An auxiliary or shutdown circulator is installed at the bottom of a reactor vessel, and it is an auxiliary circulator for shutting down a reactor in case of refueling or accelerating cooling down in case of fast cooling. Since a rotating shaft of a helium circulator is supported by bearings, bearings are the important machine elements which determines reliability of a helium circulator and a nuclear reactor. Various types of support bearings have been developed and applied for circulator bearings since 1960s, and it is still developing for developing VHTRs. So it is necessary to review and analyze the current technical state of helium circulator support bearings to develop bearings for Koran developing VHTR helium circulator

Creep-fatigue of High Temperature Materials for VHTR: Effect of Cyclic Loading and Environment

Energy Technology Data Exchange (ETDEWEB)

Celine Cabet; L. Carroll; R. Wright; R. Madland

2011-05-01

Alloy 617 is the one of the leading candidate materials for Intermediate Heat eXchangers (IHX) of a Very High Temperature Reactor (VHTR). System start-ups and shut-downs as well as power transients will produce low cycle fatigue (LCF) loadings of components. Furthermore, the anticipated IHX operating temperature, up to 950°C, is in the range of creep so that creep-fatigue interaction, which can significantly increase the fatigue crack growth, may be one of the primary IHX damage modes. To address the needs for Alloy 617 codification and licensing, a significant creep-fatigue testing program is underway at Idaho National Laboratory. Strain controlled LCF tests including hold times up to 1800s at maximum tensile strain were conducted at total strain range of 0.3% and 0.6% in air at 950°C. Creep-fatigue testing was also performed in a simulated VHTR impure helium coolant for selected experimental conditions. The creep-fatigue tests resulted in failure times up to 1000 hrs. Fatigue resistance was significantly decreased when a hold time was added at peak stress and when the total strain was increased. The fracture mode also changed from transgranular to intergranular with introduction of a tensile hold. Changes in the microstructure were methodically characterized. A combined effect of temperature, cyclic and static loading and environment was evidenced in the targeted operating conditions of the IHX. This paper This paper reviews the data previously published by Carroll and co-workers in references 10 and 11 focusing on the role of inelastic strain accumulation and of oxidation in the initiation and propagation of surface fatigue cracks.

Assessment of very high-temperature reactors in process applications

International Nuclear Information System (INIS)

Spiewak, I.; Jones, J.E. Jr.; Gambill, W.R.; Fox, E.C.

1976-11-01

An overview is presented of the technical and economic feasibility for the development of a very high-temperature reactor (VHTR) and associated processes. A critical evaluation of VHTR technology for process temperatures of 1400 and 2000 0 F is made. Additionally, an assessment of potential market impact is made to determine the commercial viability of the reactor system. It is concluded that VHTR process heat in the range of 1400 to 1500 0 F is attainable with near-term technology. However, process heat in excess of 1600 0 F would require considerably more materials development. The potential for the VHTR could include a major contribution to synthetic fuel, hydrogen, steel, and fertilizer production and to systems for transport and storage of high-temperature heat. A recommended development program including projected costs is presented

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-21

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

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

2005-01-01

Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

Application of Looped Network Analysis Method to Core of Prismatic VHTR

International Nuclear Information System (INIS)

Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

2016-01-01

Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively as shown in Fig. 1. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Moreover, it is hard to cover whole cases corresponding to the various bypass gap distribution in the whole VHTR core. In order to solve this problem, in this study, the flow network analysis code, FastNet (Flow Analysis for Steady-state Network), was developed using the Looped Network Analysis Method. The applied method was validated by comparing with SNU VHTR multi-block experiment. A 3-demensional network modeling was conducted representing flow paths as flow resistances. Flow network analysis code, FastNet, was developed to evaluate the core bypass flow distribution by using looped network analysis method. Complex flow network could be solved simply by converting the non-linear momentum equation to the linearized equation. The FastNet code predicted the flow distribution of the SNU multi-block experiment accurately

An Assessment of the VHTR Safety Distance Using the Reliability Physics Model

International Nuclear Information System (INIS)

Lee, Joeun; Kim, Jintae; Jae, Moosung

2015-01-01

In Korea planning the production of hydrogen using high temperature from nuclear power is in progress. To produce hydrogen from nuclear plants, supplying temperature above 800 .deg. C is required. Therefore, Very High Temperature Reactor (VHTR) which is able to provide about 950 .deg. C is suitable. In situation of high temperature and corrosion where hydrogen might be released easily, hydrogen production facility using VHTR has a danger of explosion. Moreover explosion not only has a bad influence upon facility itself but also on VHTR. Those explosions result in unsafe situation that cause serious damage. However, In terms of thermal-hydraulics view, long distance makes low efficiency Thus, in this study, a methodology for the safety assessment of safety distance between the hydrogen production facilities and the VHTR is developed with reliability physics model. Based on the standard safety criteria which is a value of 1 x 10 -6 , the safety distance between the hydrogen production facilities and the VHTR using reliability physics model are calculated to be a value of 60m - 100m. In the future, assessment for characteristic of VHTR, the capacity to resist pressure from outside hydrogen explosion and the overpressure for the large amount of detonation volume in detail is expected to identify more precise safety distance using this reliability physics model

Uncertainty analysis methods for estimation of reliability of passive system of VHTR

International Nuclear Information System (INIS)

Han, S.J.

2012-01-01

An estimation of reliability of passive system for the probabilistic safety assessment (PSA) of a very high temperature reactor (VHTR) is under development in Korea. The essential approach of this estimation is to measure the uncertainty of the system performance under a specific accident condition. The uncertainty propagation approach according to the simulation of phenomenological models (computer codes) is adopted as a typical method to estimate the uncertainty for this purpose. This presentation introduced the uncertainty propagation and discussed the related issues focusing on the propagation object and its surrogates. To achieve a sufficient level of depth of uncertainty results, the applicability of the propagation should be carefully reviewed. For an example study, Latin-hypercube sampling (LHS) method as a direct propagation was tested for a specific accident sequence of VHTR. The reactor cavity cooling system (RCCS) developed by KAERI was considered for this example study. This is an air-cooled type passive system that has no active components for its operation. The accident sequence is a low pressure conduction cooling (LPCC) accident that is considered as a design basis accident for the safety design of VHTR. This sequence is due to a large failure of the pressure boundary of the reactor system such as a guillotine break of coolant pipe lines. The presentation discussed the obtained insights (benefit and weakness) to apply an estimation of reliability of passive system

Monte Carlo simulation of VHTR particle fuel with chord length sampling

International Nuclear Information System (INIS)

Ji, W.; Martin, W. R.

2007-01-01

The Very High Temperature Gas-Cooled Reactor (VHTR) poses a problem for neutronic analysis due to the double heterogeneity posed by the particle fuel and either the fuel compacts in the case of the prismatic block reactor or the fuel pebbles in the case of the pebble bed reactor. Direct Monte Carlo simulation has been used in recent years to analyze these VHTR configurations but is computationally challenged when space dependent phenomena are considered such as depletion or temperature feedback. As an alternative approach, we have considered chord length sampling to reduce the computational burden of the Monte Carlo simulation. We have improved on an existing method called 'limited chord length sampling' and have used it to analyze stochastic media representative of either pebble bed or prismatic VHTR fuel geometries. Based on the assumption that the PDF had an exponential form, a theoretical chord length distribution is derived and shown to be an excellent model for a wide range of packing fractions. This chord length PDF was then used to analyze a stochastic medium that was constructed using the RSA (Random Sequential Addition) algorithm and the results were compared to a benchmark Monte Carlo simulation of the actual stochastic geometry. The results are promising and suggest that the theoretical chord length PDF can be used instead of a full Monte Carlo random walk simulation in the stochastic medium, saving orders of magnitude in computational time (and memory demand) to perform the simulation. (authors)

Graphite oxidation and structural strength of graphite support column in VHTR

International Nuclear Information System (INIS)

Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

2009-01-01

The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)

VHTR-based Nuclear Hydrogen Plant Analysis for Hydrogen Production with SI, HyS, and HTSE Facilities

International Nuclear Information System (INIS)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

2016-01-01

In this paper, analyses of material and heat balances on the SI, HyS, and HTSE processes coupled to a Very High Temperature gas-cooled Reactor (VHTR) were performed. The hydrogen production efficiency including the thermal to electric energy ratio demanded from each process is found and the normalized evaluation results obtained from three processes are compared to each other. The currently technological issues to maintain the long term continuous operation of each process will be discussed at the conference site. VHTR-based nuclear hydrogen plant analysis for hydrogen production with SI, HyS, and HTSE facilities has been carried out to determine the thermal efficiency. It is evident that the thermal to electrical energy ratio demanded from each hydrogen production process is an important parameter to select the adequate process for hydrogen production. To improve the hydrogen production efficiency in the SI process coupled to the VHTR without electrical power generation, the demand of electrical energy in the SI process should be minimized by eliminating an electrodialysis step to break through the azeotrope of the HI/I_2/H_2O ternary aqueous solution

An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

International Nuclear Information System (INIS)

Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

2006-01-01

KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

An investigation of structural design methodology for HTGR reactor internals with ceramic materials (Contract research)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro

2008-03-01

To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR. (author)

Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

International Nuclear Information System (INIS)

Guzina, Bojan; Kunerth, Dennis

2014-01-01

A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

Three-dimensional NDE of VHTR core components via simulation-based testing. Final report

Energy Technology Data Exchange (ETDEWEB)

Guzina, Bojan [Univ. of Minnesota, Minneapolis, MN (United States); Kunerth, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-30

A next generation, simulation-driven-and-enabled testing platform is developed for the 3D detection and characterization of defects and damage in nuclear graphite and composite structures in Very High Temperature Reactors (VHTRs). The proposed work addresses the critical need for the development of high-fidelity Non-Destructive Examination (NDE) technologies for as-manufactured and replaceable in-service VHTR components. Centered around the novel use of elastic (sonic and ultrasonic) waves, this project deploys a robust, non-iterative inverse solution for the 3D defect reconstruction together with a non-contact, laser-based approach to the measurement of experimental waveforms in VHTR core components. In particular, this research (1) deploys three-dimensional Scanning Laser Doppler Vibrometry (3D SLDV) as a means to accurately and remotely measure 3D displacement waveforms over the accessible surface of a VHTR core component excited by mechanical vibratory source; (2) implements a powerful new inverse technique, based on the concept of Topological Sensitivity (TS), for non-iterative elastic waveform tomography of internal defects - that permits robust 3D detection, reconstruction and characterization of discrete damage (e.g. holes and fractures) in nuclear graphite from limited-aperture NDE measurements; (3) implements state-of-the art computational (finite element) model that caters for accurately simulating elastic wave propagation in 3D blocks of nuclear graphite; (4) integrates the SLDV testing methodology with the TS imaging algorithm into a non-contact, high-fidelity NDE platform for the 3D reconstruction and characterization of defects and damage in VHTR core components; and (5) applies the proposed methodology to VHTR core component samples (both two- and three-dimensional) with a priori induced, discrete damage in the form of holes and fractures. Overall, the newly established SLDV-TS testing platform represents a next-generation NDE tool that surpasses

Designing a TAC thermometer from a VHTR graphite structure

Energy Technology Data Exchange (ETDEWEB)

Smith, James A., E-mail: James.Smith@INL.gov; Kotter, Dale, E-mail: James.Smith@INL.gov [Fuel Performance and Design, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garrett, Steven L.; Ali, Randall A. [Graduate Program in Acoustics, Penn State University, State College, PA (United States)

2015-03-31

The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. Very High Temperature Reactors are pushing the in core temperatures even higher. A unique sensing approach will be discussed to address the necessary high temperature measurements. Thermoacoustic thermometry exploits high temperatures and uses materials that are immune to the effects of ionizing radiation to create a temperature sensor that is self-powered and wireless. In addition, the form-factor for the Thermoacoustic Thermometer (TACT) can be designed to be integrated within common in-pile structures. There are no physical moving parts required for TACT and the sensor is self-powered, as it uses the nuclear fuel for its heat source. TACT data will be presented from a laboratory prototype mimicking the design necessary for a VHTR graphite structure.

Improvements of Physical Models in TRITGO code for Tritium Behavior Analysis in VHTR

International Nuclear Information System (INIS)

Yoo, Jun Soo; Tak, Nam Il; Lim, Hong Sik

2010-01-01

Since tritium is radioactive material with 12.32 year of half-life and is generated by a ternary fission reaction in fuel as well as by neutron absorption reactions of impurities in Very High Temperature gas-cooled Reactor (VHTR) core, accurate prediction of tritium behavior and its concentration in product hydrogen is definitely important in terms of public safety for its construction. In this respect, TRITGO code was developed for estimating the tritium production and distribution in high temperature gas-cooled reactors by General Atomics (GA). However, some models in it are hard-wired to specific reactor type or too simplified, which makes the analysis results less applicable. Thus, major improvements need to be considered for better predictions. In this study, some of model improvements have been suggested and its effect is evaluated based on the analysis work against PMR600 design concept

Evaluation of Core Bypass Flow in the Prismatic VHTR with a Multi-block Experiment

International Nuclear Information System (INIS)

Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Kim, Min Hwan

2010-01-01

The core of Prismatic Modular Reactor (PMR) consists of assemblies of hexagonal graphite fuel and reflector elements. The core bypass flow of Very High Temperature Reactor (VHTR) is defined as the core flow that does not pass through the coolant channels but passes through the bypass gap between fuel elements. The increase in bypass flow makes the decrease in effective coolant flow. Since the core bypass flow has a negative impact on safety and efficiency of VHTR, core bypass phenomena have to be investigated to improve the core thermal margin of VHTR. For this purpose, the international project, I-NERI project, has been carried out since 2008. I-NERI project is collaborative project that KAERI and SNU of Korea side and INL, ANL and TAMU of U.S side are involved. In order to evaluate the core bypass flow, the multicolumn and multi-layer experimental facility is designed by SNU. In this experiment, the effect of cross-flow and local variation of bypass gap on the bypass flow distribution is investigated. Furthermore, the experimental data will be used for validation of CFD code or thermal hydraulic analysis codes such as GAMMA or GAS-NET

Power requirements at the VHTR/HTE interface for hydrogen production

International Nuclear Information System (INIS)

Vilim, R.B.

2007-01-01

The power requirements at the interface between the High Temperature Electrolysis (HTE) process and the Very High Temperature Reactor (VHTR) were investigated. The study was performed using a network systems code that linked together individual component models for boiler, condenser, turbine, compressor, pump, gas-to-gas heat exchanger, electrolyser, and reactor and properties for water, hydrogen, oxygen, nitrogen, and helium. A species mixture model supported the use of mixtures of gases in each component model. The requirements for a reference design with a dedicated high temperature process heat loop are given. In general the quantity and quality of the process heat needed by the HTE process is a function of how the electrolyser is operated. Operating at higher voltage increases throughput and resistive heating providing the opportunity to recuperate this heat and supplant a large fraction of high temperature reactor heat. Any shortfall can be added by electrical heaters in the HTE plant. Eliminating the associated high temperature heat exchanger from the nuclear plant in this manner would significantly improve safety and maintainability. Low temperature process heat is still needed to vaporize water for the HTE process but this can be obtained at very low cost from VHTR waste heat rejected to the ultimate heat sink. (author)

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

Energy Technology Data Exchange (ETDEWEB)

Moses, David Lewis [ORNL

2011-10-01

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

Study on the tritium behaviors in the VHTR system. Part 2: Analyses on the tritium behaviors in the VHTR/HTSE system

Energy Technology Data Exchange (ETDEWEB)

Kim, Eung S. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Oh, Chang H., E-mail: Chang.Oh@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States); Patterson, Mike [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3885 (United States)

2010-07-15

Tritium behaviors in the very high temperature gas reactor (VHTR)/high temperature steam electrolysis (HTSE) system have been analyzed by the TPAC developed by Idaho National Laboratory (INL). The reference system design and conditions were based on the indirect parallel configuration between a VHTR and a HTSE. The analyses were based on the SOBOL method, a modern uncertainty and sensitivity analyses method using variance decomposition and Monte Carlo method. A total of 14 parameters have been taken into account associated with tritium sources, heat exchangers, purification systems, and temperatures. Two sensitivity indices (first order index and total index) were considered, and 15,360 samples were totally used for solution convergence. As a result, important parameters that affect tritium concentration in the hydrogen product have been identified and quantified with the rankings. Several guidelines and recommendations for reducing modeling uncertainties have been also provided throughout the discussions along with some useful ideas for mitigating tritium contaminations in the hydrogen product.

Computational Fluid Dynamic Analysis of the VHTR Lower Plenum Standard Problem

International Nuclear Information System (INIS)

Johnson, Richard W.; Schultz, Richard R.

2009-01-01

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U.S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made

Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

2007-07-15

Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

International Nuclear Information System (INIS)

Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok

2016-01-01

The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

Energy Technology Data Exchange (ETDEWEB)

Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

2006-05-15

For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

International Nuclear Information System (INIS)

Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

2006-05-01

For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels

Feasibility study on the application of carbide (ZrC, SiC) for VHTR

Energy Technology Data Exchange (ETDEWEB)

Park, Ji Yeon; Kim, Weon Ju; Jung, Choong Hwan; Ryu, Woo Seog; Kim, Si Hyeong; Jang, Moon Hee; Lee, Young Woo

2006-08-15

A feasibility study on the coating process of ZrC for the TRISO nuclear fuel and applications of SiC as high temperature materials for the core components has performed to develop the fabrication process for the advanced ZrC TRISO fuels and the high temperature structural components for VHTR, respectively. In the case of ZrC coating, studies were focused on the comparisons of the developed coating processes for screening of our technology, the evaluations of the reactions parameters for a ZrC deposition by the thermodynamic calculations and the preliminary coating experiments by the chloride process. With relate to SiC ceramics, our interesting items are as followings; an analysis of applications and specifications of the SiC components and collections of the SiC properties and establishments of data base. For these purposes, applications of SiC ceramics for the GEN-IV related components as well as the fusion reactor related ones were reviewed. Additionally, the on-going activities with related to the ZrC clad and the SiC composites discussed in the VHTR GIF-PMB, were reviewed to make the further research plans at the section 1 in chapter 3.

Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

Energy Technology Data Exchange (ETDEWEB)

Glenn E McCreery; Keith G Condie

2006-09-01

The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

International Nuclear Information System (INIS)

Marvin, M.D.

1978-01-01

Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR and PP)

International Nuclear Information System (INIS)

Moses, David Lewis

2011-01-01

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR and PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR and PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR and PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR and PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet

Design studies for the Mark-III core of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

1979-08-01

The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.

1976-12-01

In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics

International Nuclear Information System (INIS)

Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z.

2013-01-01

This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR

Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

Energy Technology Data Exchange (ETDEWEB)

Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

2014-09-15

Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

Feasibility study of thermal insulation materials for core support of experimental VHTR

International Nuclear Information System (INIS)

Kawakami, H.; Nakanishi, T.

1982-01-01

Thermal insulation materials for core support of the experimental VHTR, planned by JAERI, should maintain moderate compressive strength and dimensional stability as well as low thermal conductivity at the maximum service temperature of 1100 0 C for 20 years. For selecting materials, we investigate properties of some candidates, and evaluate their feasibility. Preliminary tests, heat treatment test and compressive creep tests for 1000 hours at 900 0 C and 1000 0 C were conducted. In the preliminary tests, EG-38B (carbon baked at 1350 0 C) and Fine Finnex 600 (silicon nitride) showed acceptable physical stability. In the heat treatment tests, silicon nitride showed weight loss probably caused by thermal decomposition. Compressive creep deformation of Fine Finnex 600 was negligible under stress of 100 kg/cm 2 for 1000 hours. Heat treatment at 1200 to 1300 0 C for 50 hours improved dimensional stability of carbon at 1000 0 C

A study on a reliability assessment methodology for the VHTR safety systems

International Nuclear Information System (INIS)

Lee, Hyung Sok

2012-02-01

The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

Current status of VHTR development in Japan

International Nuclear Information System (INIS)

Aochi, A.; Kondo, T.

1982-01-01

The status of the program at the beginning of fiscal 1982 is reviewed. Special emphasis is placed on the altering of the output helium temperature of the experimental VHTR to 950 0 . The modification is aimed at establishing the technical basis for post-experimental VHTR output helium temperature of 1000 0 C. Notes are given on the design of the VHTR as well as various research and development efforts in Japan on multi-purpose nuclear heat applications and HTGR technology

FY-05 Second Quarter Report On Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving PBR Efficiency and Testing Material Compatibility

International Nuclear Information System (INIS)

Chang Oh

2005-01-01

The objective of this research is to improve a helium Brayton cycle and to develop a supercritical carbon dioxide Brayton cycle for the Pebble Bed Reactor (PBR) that can also be applied to the Fast Gas-Cooled Reactor (FGR) and the Very-High-Temperature Gas-Cooled Reactor (VHTR). The proposed supercritical carbon dioxide Brayton cycle will be used to improve the PBR, FGR, and VHTR net plant efficiency. Another objective of this research is to test materials to be used in the power conversion side at supercritical carbon dioxide conditions. Generally, the optimized Brayton cycle and balance of plant (BOP) to be developed from this study can be applied to Generation-IV reactor concepts. Particularly, we are interested in VHTR because it has a good chance of being built in the near future

Design studies of back up cores for the experimental multi-purpose VHTR, (1)

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

1982-09-01

For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

International Nuclear Information System (INIS)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

2015-01-01

Highlights: • A 1/8th geometric-scale test facility that models the VHTR hot plenum is proposed. • Geometric scaling analysis is introduced for VHTR to analyze air-ingress accident. • Design calculations are performed to show that accident phenomenology is preserved. • Some analyses include time scale, hydraulic similarity and power scaling analysis. • Test facility has been constructed and shake-down tests are currently being carried out. - Abstract: A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time

Numerical investigation of the High Temperature Reactor (VHTR) using computational fluid dynamics; Investigacao numerica do Reator de Alta Temperatura (VHTR) utilizando fluidodinamica computacional

Energy Technology Data Exchange (ETDEWEB)

Pinto, Joao Pedro C.T.A.; Santos, Andre A. Campagnole dos; Mesquita, Amir Z., E-mail: jpctap@cdtn.br, E-mail: aacs@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG),Belo Horizonte, MG (Brazil). Lab. de Termo-Hidraulica

2013-07-01

This work consists to evaluate and continue the study that is being developed in the Laboratory of Thermo-Hydraulics of the CNEN/CDTN (Centro de Desenvolvimento da Tecnologia Nuclear), aiming to validate the methods and procedures used in the numerical calculations of fluid flow in fuel elements of the core of the VHTR.

Safety study of the coupling of a VHTR with a hydrogen production plant

International Nuclear Information System (INIS)

Bertrand, F.; Germain, T.; Bentivoglio, F.; Bonnet, F.; Moyart, Q.; Aujollet, P.

2011-01-01

Highlights: → The paper deals with safety issues of the coupling of a VHTR with a H 2 production plant. → Internal incidents/accidents in the coupling system have been studied with the CATHARE2 code. → Transient studies have indicated a substantial grace delay when the VHTR faces the H 2 plant disturbances. → Hydrogen release and combustion leads to safety distances of about 100 m. → No showstopper has been put in evidence regarding the feasibility of the VHTR/H 2 plant coupling. - Abstract: The present paper deals with specific safety issues resulting from the coupling of a nuclear reactor (very high temperature reactor, VHTR) with a hydrogen production plant (HYPP). The first part is devoted to the safety approach consisting in taking into account the safety standards and rules dedicated to the nuclear facility as well as those dedicated to the process industry. This approach enabled two main families of events to be distinguished: the so-called internal events taking place in the coupling circuit (transients, breaks in pipes and in heat exchangers) and the external events able to threat the integrity of the various equipments (in particular the VHTR containment and emergency cooling system) that could result from accidents in the HYPP. By considering a hydrogen production by means of the iodine/sulfur (IS) process, the consequences of the both families of events aforementioned have been assessed in order to provide an order of magnitude of the effects of the incidents and accidents and also in order to propose safety provisions to mitigate these effects when it is necessary. The study of transients induced by a failure of a part of the HYPP has shown the possibility to keep the part of the HYPP unaffected by the transient under operation by means of an adapted regulation set. Moreover, the time to react in case of transfer of corrosive products in the VHTR containment has been assessed as well as the thermohydraulic loading that would experience the

Validation Studies for Numerical Simulations of Flow Phenomena Expected in the Lower Plenum of a Prismatic VHTR Reference Design

International Nuclear Information System (INIS)

Richard W. Johnson

2005-01-01

The final design of the very high temperature reactor (VHTR) of the fourth generation of nuclear power plants (Gen IV) has not yet been established. The VHTR may be either a prismatic (block) or pebble bed type. It may be either gas-cooled or cooled with an as yet unspecified molten salt. However, a conceptual design of a gas-cooled VHTR, based on the General Atomics GT-MHR, does exist and is called the prismatic VHTR reference design, MacDonald et al [2003], General Atomics [1996]. The present validation studies are based on the prismatic VHTR reference design. In the prismatic VHTR reference design, the flow in the lower plenum will be introduced by dozens of turbulent jets issuing into a large crossflow that must negotiate dozens of cylindrical support columns as it flows toward the exit duct of the reactor vessel. The jets will not all be at the same temperature due to the radial variation of power density expected in the core. However, it is important that the coolant be well mixed when it enters the power conversion unit to ensure proper operation and long life of the power conversion machinery. Hence, it is deemed important to be able to accurately model the flow and mixing of the variable temperature coolant in the lower plenum and exit duct. Accurate flow modeling involves determining modeling strategies including the fineness of the grid needed, iterative convergence tolerance, numerical discretization method used, whether the flow is steady or unsteady, and the turbulence model and wall treatment employed. It also involves validation of the computer code and turbulence model against a series of separate and combined flow phenomena and selection of the data used for the validation. The present report describes progress made to date for the task entitled ''CFD software validation of jets in crossflow'' which was designed to investigate the issues pertaining to the validation process

Effect of Permanent Side Reflector on the Temperature Variation in the VHTR Core

Energy Technology Data Exchange (ETDEWEB)

Lee, Sung Nam; Tak, Nam-il; Kim, Min-Hwan [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The temperature and pressure conditions range from 490°C to 950°C, 7MPa. GAMMA+ was developed to predict the overall phenomena of the VHTR system. The GAMMA+ algorithms focused on the transient condition for the systems. Therefore, the computational control volumes are coarse for reducing the computational time. However, there are difficulties calculating the temperature gradient in the fuel blocks in detail. There is a demand to predict a hot spot and temperature distribution in the reactor core to apply a thermal stress and find the fuel temperature margin. Computational Fluid Dynamic (CFD) tools can be an option to model the VHTR. However, the fluid has to be solved in three dimensions. The long computational time and heavy burden of the memory size have called for an alternative option. The PSR blocks are considered in the prismatic VHTR calculation with the CORONA code. The temperatures of a single assembly with an arc shape reflector by the CORONA code were verified with the results by the CFX calculation. The temperature distributions of the PSR regions did not show significant differences depending on the fixed inlet temperature boundary condition and bypass flow condition.

NERI Quarterly Progress Report -- April 1 - June 30, 2005 -- Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving PBR Efficiency and Testing Material Compatibility

International Nuclear Information System (INIS)

Chang Oh

2005-01-01

The objective of this research is to improve a helium Brayton cycle and to develop a supercritical carbon dioxide Brayton cycle for the Pebble Bed Reactor (PBR) that can also be applied to the Fast Gas-Cooled Reactor (FGR) and the Very-High-Temperature Gas-Cooled Reactor (VHTR). The proposed supercritical carbon dioxide Brayton cycle will be used to improve the PBR, FGR, and VHTR net plant efficiency. Another objective of this research is to test materials to be used in the power conversion side at supercritical carbon dioxide conditions. Generally, the optimized Brayton cycle and balance of plant (BOP) to be developed from this study can be applied to Generation-IV reactor concepts. Particularly, we are interested in VHTR because it has a good chance of being built in the near future

Structural materials performance research at JRC-Institute for Energy

International Nuclear Information System (INIS)

Haehner, P.

2009-01-01

The DG-JRC structure and activities are presented in the paper. The Generation IV reactor concepts Very High Temperature Reactor (VHTR), Supercritical Water Reactor (SCWR) and Lead Cooled Reactor (LCR) are currently under study at the JRC. Requirements for innovative nuclear systems and material-related operational condition are under investigation. Considering the operational experience with current nuclear industry, these conditions imply demanding challenges from the structural materials point of view. The European Projects and initiatives and coordinated research programs are also presented

Critical evaluation of high-temperature gas-cooled reactors applicable to coal conversion

International Nuclear Information System (INIS)

Spiewak, I.; Jones, J.E. Jr.; Rittenhouse, P.L.; DeStefano, J.R.; Delene, J.G.

1975-12-01

A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 2000 0 F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 1500 0 F could be developed with a high degree of assurance. Process heat at 1600 0 F would require considerably more materials development. While temperatures up to 2000 0 F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

International Nuclear Information System (INIS)

Hawari, Ayman

2014-01-01

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

High Temperature Materials Interim Data Qualification Report

International Nuclear Information System (INIS)

Lybeck, Nancy

2010-01-01

Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim FY2010 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under NQA-1 guidelines, and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from two test series within the High Temperature Materials data stream have been entered into the NDMAS vault: (1) Tensile Tests for Sm (i.e., Allowable Stress) Confirmatory Testing - 1,403,994 records have been inserted into the NDMAS database. Capture testing is in process. (2) Creep-Fatigue Testing to Support Determination of Creep-Fatigue Interaction Diagram - 918,854 records have been processed and inserted into the NDMAS database. Capture testing is in process.

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

2015-04-29

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

International Nuclear Information System (INIS)

Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Park, Jong Woon

2011-01-01

The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

CFD Validation with a Multi-Block Experiment to Evaluate the Core Bypass Flow in VHTR

International Nuclear Information System (INIS)

Yoon, Su Jong; Lee, Jeong Hun; Park, Goon Cherl; Kim, Min Hwan

2010-01-01

Core bypass flow of Very High Temperature Reactor (VHTR) is defined as the ineffective coolant which passes through the bypass gaps between the block columns and the crossflow gaps between the stacked blocks. This flows lead to the variation of the flow distribution in the core and affect the core thermal margin and the safety of VHTR. Therefore, bypass flow should be investigated and quantified. However, it is not a simple question, because the flow path of VHTR core is very complex. In particular, since dimensions of the bypass gap and the crossflow gap are of the order of few millimeters, it is very difficult to measure and to analyze the flow field at those gaps. Seoul National University (SNU) multi-block experiment was carried out to evaluate the bypass flow distribution and the flow characteristics. The coolant flow rate through outlet of each block column was measured, but the local flow field was measured restrictively in the experiment. Instead, CFD analysis was carried out to investigate the local phenomena of the experiment. A commercial CFD code CFX-12 was validated by comparing the simulation results and the experimental data

Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

Energy Technology Data Exchange (ETDEWEB)

Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

2005-12-01

The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

International Nuclear Information System (INIS)

Sato, Hiroyuki; Johnson, Richard; Schultz, Richard

2010-01-01

The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

Summary of the experimental multi-purpose very high temperature gas cooled reactor design

International Nuclear Information System (INIS)

1984-12-01

The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

International Nuclear Information System (INIS)

1980-01-01

Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050 0 C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850 0 C on several experimental alloys are discussed

Research and development for high temperature gas cooled reactor in Japan

International Nuclear Information System (INIS)

Taketani, K.

1978-01-01

The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

International Nuclear Information System (INIS)

Casino, William A. Jr.

2006-01-01

The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

TRISO fuel thermal simulations in the LS-VHTR

Energy Technology Data Exchange (ETDEWEB)

Ramos, Mario C.; Scari, Maria E.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F., E-mail: marc5663@gmail.com, E-mail: melizabethscari@yahoo.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@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

The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li{sub 2}BeF{sub 4} called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed. (author)

High Temperature Materials Interim Data Qualification Report FY 2011

International Nuclear Information System (INIS)

Lybeck, Nancy

2011-01-01

Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim fiscal year (FY) 2011 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under the Nuclear Quality Assurance (NQA)-1 guidelines and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from seven test series within the High Temperature Materials data stream have been entered into the NDMAS vault, including tensile tests, creep tests, and cyclic tests. Of the 5,603,682 records currently in the vault, 4,480,444 have been capture passed, and capture testing is in process for the remaining 1,123,238.

Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

International Nuclear Information System (INIS)

Steven R. Sherman

2007-01-01

The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant

Utilization of Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs, Advantages and Limitations

International Nuclear Information System (INIS)

Tsvetkov, Pavel V.

2009-01-01

This project assessed the advantages and limitations of using minor actinides as a fuel component to achieve ultra-long life Very High Temperature Reactor (VHTR) configurations. Researchers considered and compared the capabilities of pebble-bed and prismatic core designs with advanced actinide fuels to achieve ultra-long operation without refueling. Since both core designs permit flexibility in component configuration, fuel utilization, and fuel management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. The project studied advanced actinide fuels, which could reduce the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository and enable recovery of the energy contained in spent fuel. The ultra-long core life autonomous approach may reduce the technical need for additional repositories and is capable to improve marketability of the Generation IV VHTR by allowing worldwide deployment, including remote regions and regions with limited industrial resources. Utilization of minor actinides in nuclear reactors facilitates developments of new fuel cycles towards sustainable nuclear energy scenarios.

Aqueous alteration of VHTR fuels particles under simulated geological conditions

Energy Technology Data Exchange (ETDEWEB)

Ait Chaou, Abdelouahed, E-mail: aitchaou@subatech.in2p3.fr; Abdelouas, Abdesselam; Karakurt, Gökhan; Grambow, Bernd

2014-05-01

Very High Temperature Reactor (VHTR) fuels consist of the bistructural-isotropic (BISO) or tristructural-isotropic (TRISO)-coated particles embedded in a graphite matrix. Management of the spent fuel generated during VHTR operation would most likely be through deep geological disposal. In this framework we investigated the alteration of BISO (with pyrolytic carbon) and TRISO (with SiC) particles under geological conditions simulated by temperatures of 50 and 90 °C and in the presence of synthetic groundwater. Solid state (scanning electron microscopy (SEM), micro-Raman spectroscopy, electron probe microanalyses (EPMA) and X-ray photoelectron spectroscopy (XPS)) and solution analyses (ICP-MS, ionique chromatography (IC)) showed oxidation of both pyrolytic carbon and SiC at 90 °C. Under air this led to the formation of SiO{sub 2} and a clay-like Mg–silicate, while under reducing conditions (H{sub 2}/N{sub 2} atmosphere) SiC and pyrolytic carbon were highly stable after a few months of alteration. At 50 °C, in the presence and absence of air, the alteration of the coatings was minor. In conclusion, due to their high stability in reducing conditions, HTR fuel disposal in reducing deep geological environments may constitute a viable solution for their long-term management.

WAVE PROPAGATION in the HOT DUCT of VHTR

Energy Technology Data Exchange (ETDEWEB)

Richard Schultz; Jim C. P. Liou

2013-07-01

In VHTR, helium from the reactor vessel is conveyed to a power conversion unit through a hot duct. In a hypothesized Depressurized Conduction Cooldown event where a rupture of the hot duct occurs, pressure waves will be initiated and reverberate in the hot duct. A numerical model is developed to quantify the transients and the helium mass flux through the rupture for such events. The flow path of the helium forms a closed loop but only the hot duct is modeled in this study. The lower plum of the reactor vessel and the steam generator are treated as specified pressure and/or temperature boundary to the hot duct. The model is based on the conservation principles of mass, momentum and energy, and on the equations of state for helium. The numerical solution is based on the method of characteristics with specified time intervals with a predictor and corrector algorithm. The rupture sub-model gives reasonable results. Transients induced by ruptures with break area equaling 20%, 10%, and 5% of the duct cross-sectional area are described.

An Innovative VHTR Waste Heat Integration with Forward Osmosis Desalination Process

Energy Technology Data Exchange (ETDEWEB)

Park, Min Young; Kim, Eung Soo [Seoul National Univ., Seoul (Korea, Republic of)

2013-10-15

The integration concept implies the coupling of the waste heat from VHTR with the draw solute recovery system of FO process. By integrating these two novel technologies, advantages, such as improvement of total energy utilization, and production of fresh water using waste heat, can be achieved. In order to thermodynamically analyze the integrated system, the FO process and power conversion system of VHTR are simulated using chemical process software UNISIM together with OLI property package. In this study, the thermodynamic analysis on the VHTR and FO integrated system has been carried out to assess the feasibility of the concept. The FO process including draw solute recovery system is calculated to have a higher GOR compared to the MSF and MED when reasonable FO performance can be promised. Furthermore, when FO process is integrated with the VHTR to produce potable water from waste heat, it still shows a comparable GOR to typical GOR values of MSF and MED. And the waste heat utilization is significantly higher in FO than in MED and MSF. This results in much higher water production when integrated to the same VHTR plant. Therefore, it can be concluded that the suggested integrated system of VHTR and FO is a very promising and strong system concept which has a number of advantages over conventional technologies.

Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

International Nuclear Information System (INIS)

Chang Oh

2006-01-01

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

Advanced construction materials for thermo-chemical hydrogen production from VHTR process heat

International Nuclear Information System (INIS)

Kosmidou, Theodora; Haehner, Peter

2009-01-01

The (very) high temperature reactor concept ((V)HTR) is characterized by its potential for process heat applications. The production of hydrogen by means of thermo-chemical cycles is an appealing example, since it is more efficient than electrolysis due to the direct use of process heat. The sulfur-iodine cycle is one of the best studied processes for the production of hydrogen, and solar or nuclear energy can be used as a heating source for the high temperature reaction of this process. The chemical reactions involved in the cycle are: I 2 (l) + SO 2 (g) +2 H 2 O (l) → 2HI (l) + H 2 SO 4 (l) (70-120 deg. C); H 2 SO 4 (l) → H 2 O (l) + SO 2 (g) + 1/2 O 2 (g) (800-900 deg. C); 2HI (l) → I 2 (g) + H 2 (g) (300-450 deg. C) The high temperature decomposition of sulphuric acid, which is the most endothermic reaction, results in a very aggressive chemical environment which is why suitable materials for the decomposer heat exchanger have to be identified. The class of candidate materials for the decomposer is based on SiC. In the current study, SiC based materials were tested in order to determine the residual mechanical properties (flexural strength and bending modulus, interfacial strength of brazed joints), after exposure to an SO 2 rich environment, simulating the conditions in the hydrogen production plant. Brazed SiC specimens were tested after 20, 100, 500 and 1000 hrs exposure to SO 2 rich environment at 850 o C under atmospheric pressure. The gas composition in the corrosion rig was: 9.9 H 2 O, 12.25 SO 2 , 6.13 O 2 , balance N 2 (% mol). The characterization involved: weight change monitoring, SEM microstructural analysis and four-point bending tests after exposure. Most of the specimens gained weight due to the formation of a corrosion layer as observed in the SEM. The corrosion treatment also showed an effect on the mechanical properties. In the four-point bending tests performed at room temperature and at 850 deg. C, a decrease in bending modulus with

High Temperature Degradation Behavior and its Mechanical Properties of Inconel 617 alloy for Intermediate Heat Exchanger of VHTR

International Nuclear Information System (INIS)

Jo, Tae Sun; Kim, Se Hoon; Kim, Young Do; Park, Ji Yeon

2008-01-01

Inconel 617 alloy is a candidate material of intermediate heat exchanger (IHX) and hot gas duct (HGD) for very high temperature reactor (VHTR) because of its excellent strength, creep-rupture strength, stability and oxidation resistance at high temperature. Among the alloying elements in Inconel 617, chromium (Cr) and aluminum (Al) can form dense oxide that act as a protective surface layer against degradation. This alloy supports severe operating conditions of pressure over 8 MPa and 950 .deg. C in He gas with some impurities. Thus, high temperature stability of Inconel 617 is very important. In this work, the oxidation behavior of Inconel 617 alloy was studied by exposure at high temperature and was discussed the high temperature degradation behavior with microstructural changes during the surface oxidation

Experimental facilities for Generation IV reactors research

International Nuclear Information System (INIS)

Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

2013-06-01

Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

Numerical investigation of flow characteristics in a prototypical lower plenum of a prismatic VHTR

International Nuclear Information System (INIS)

Ying, Alice; Narula, Manmeet; Abdou, Mohamed; Tsai, Peter; Ando, Yuya

2007-01-01

The aim of this study is to obtain insights into the flow behavior, as well as to develop predictive capability with regards to the flow and thermal mixing, that occurs in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In this paper, numerical modeling has been used to capture qualitative phenomena observed during an experiment performed at INL, using a finite volume, thermo-fluid solver system, 'SC/Tetra' from CRADLE. The choice of the correct turbulence model is critical to accurately predict the flow in the VHTR lower plenum. Four different turbulence models have been used in this study and the flow predictions are significantly different. A trail of marker particles and fluid temperature as a passive scalar have been used to qualitatively study the flow characteristics, specifically the turbulent mixing of water jets. The quantitative experimental data, when available, will be used to compare and improve on the available turbulence models. Preliminary numerical modeling has been carried out to address the issue of hot streaking and buoyancy effects of hot helium jets in the lower plenum. (author)

Experimental and computational studies of thermal mixing in next generation nuclear reactors

Science.gov (United States)

Landfried, Douglas Tyler

The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

Scaled Experimental Modeling of VHTR Plenum Flows

Energy Technology Data Exchange (ETDEWEB)

ICONE 15

2007-04-01

Abstract The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. Various scaled heated gas and water flow facilities were investigated for modeling VHTR upper and lower plenum flows during the decay heat portion of a pressurized conduction-cooldown scenario and for modeling thermal mixing and stratification (“thermal striping”) in the lower plenum during normal operation. It was concluded, based on phenomena scaling and instrumentation and other practical considerations, that a heated water flow scale model facility is preferable to a heated gas flow facility and to unheated facilities which use fluids with ranges of density to simulate the density effect of heating. For a heated water flow lower plenum model, both the Richardson numbers and Reynolds numbers may be approximately matched for conduction-cooldown natural circulation conditions. Thermal mixing during normal operation may be simulated but at lower, but still fully turbulent, Reynolds numbers than in the prototype. Natural circulation flows in the upper plenum may also be simulated in a separate heated water flow facility that uses the same plumbing as the lower plenum model. However, Reynolds number scaling distortions will occur at matching Richardson numbers due primarily to the necessity of using a reduced number of channels connected to the plenum than in the prototype (which has approximately 11,000 core channels connected to the upper plenum) in an otherwise geometrically scaled model. Experiments conducted in either or both facilities will meet the objectives of providing benchmark data for the validation of codes proposed for NGNP designs and safety studies, as well as providing a better understanding of the complex flow phenomena in the plenums.

Optimized core design and fuel management of a pebble-bed type nuclear reactor

NARCIS (Netherlands)

Boer, B.

2009-01-01

The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

Energy Technology Data Exchange (ETDEWEB)

Chang Oh

2006-03-01

The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

2016-10-15

A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

Study on Heat Transfer Characteristics of One Side Heated Vertical Channel Applied as Vessel Cooling System

International Nuclear Information System (INIS)

Kuriyama, Shinji; Takeda, Tetsuaki; Funatani, Shumpei

2014-01-01

The inherent properties of the Very-High-Temperature Reactor facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However; it is still not clear if the VHTR can maintain a passive safe function during the severe accident, or what would be a design criterion to guarantee the VHTR with the high degree of passive safe performances during the accidents. In the Very High Temperature Reactor (VHTR) which is a next generation nuclear reactor system, ceramics and graphite are used as a fuel coating material and a core structural material, respectively. Even if the depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). This study is to develop the passive cooling system for the VHTR using the vertical channel inserting porous materials. The objective of this study is to investigate heat transfer characteristics of natural convection of a one-side heated vertical channel inserting the porous materials with high porosity. In order to obtain the heat transfer and fluid flow characteristics of a vertical channel inserting porous material, we have also carried out a numerical analysis using the commercial CFD code. From the analytical results obtained in the natural convection cooling, an amount of removed heat enhanced inserting the copper wire. It was found that an amount of removed heat inserting the copper wire (porosity = 0.9972) was about 10% higher than that without the copper wire. This paper describes a thermal performance of the one-side heated vertical channel inserting copper wire with high porosity. (author)

A Design of He-Molten Salt Intermediate Heat Exchanger for VHTR

International Nuclear Information System (INIS)

Jeong, Hui Seong; Bang, Kwang Hyun

2010-01-01

The Very High Temperature Reactor (VHTR), one of the most challenging next generation nuclear reactors, has recently drawn an international interest due to its higher efficiency and the operating conditions adequate for supplying process heat to the hydrogen production facilities. To make the design of VHTR complete and plausible, the designs of the Intermediate Heat Transport Loop (IHTL) as well as the Intermediate Heat Exchanger (IHX) are known to be one of the difficult engineering tasks due to its high temperature operating condition (up to 950 .deg. C). A type of compact heat exchangers such as printed circuit heat exchanger (PCHE) has been recommended for the IHX in the technical and economical respects. Selection of the heat transporting fluid for the intermediate heat transport loop is important in consideration of safety and economical aspects. Although helium is currently the primary interest for the intermediate loop fluid, several safety concerns of gas fluids have been expressed. If the pressure boundary of the intermediate loop fails, the blowdown of the gas may overcool the reactor core and then the heat sink is lost after the blowdown. Also the large inventory of gas in the intermediate loop may leak into the primary side. There is also a recommendation that the nuclear plant and the hydrogen production plant be separated by a certain distance to ensure the safety of the nuclear plant in case of accidental heavy gas release from the chemical plant. In this circumstance, the pumping power of gas fluid in the intermediate loop will be large enough to degrade the economics of nuclear hydrogen.An alternative candidate for the intermediate loop fluid in consideration of these safety and economical problems of gas fluid can be molten salts. The Flinak molten salt, a eutectic mixture of LiF, NaF and KF (46.5:11.5:42.0 mole %) is considered to be a potential candidate for the heat transporting fluid in the IHTL due to its chemical stability against the

Development of Very High Temperature Reactor Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, Y. H.

2009-04-01

For an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

Next Generation Nuclear Plant Materials Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

G.O. Hayner; R.L. Bratton; R.N. Wright

2005-09-01

The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for

Corrosion of nickel-base heat resistant alloys in simulated VHTR coolant helium at very high temperatures

International Nuclear Information System (INIS)

Shindo, Masami; Kondo, Tatsuo

1976-01-01

A comparative evaluation was made on three commercial nickel-base heat resistant alloys exposed to helium-base atmosphere at 1000 0 C, which contained several impurities in simulating the helium cooled very high temperature nuclear reactor (VHTR) environment. The choice of alloys was made so that the effect of elements commonly found in commercial alloys were typically examined. The corrosion in helium at 1000 0 C was characterized by the sharp selection of thermodynamically unstable elements in the oxidizing process and the resultant intergranular penetration and internal oxidation. Ni-Cr-Mo-W type solution hardened alloy such as Hastelloy-X showed comparatively good resistance. The alloy containing Al and Ti such as Inconel-617 suffered adverse effect in contrast to its good resistance to air oxidation. The alloy nominally composed only of noble elements, Ni, Fe and Mo, such as Hastelloy-B showed least apparent corrosion, while suffered internal oxidation due to small amount of active impurities commonly existing in commercial heats. The results were discussed in terms of selection and improvement of alloys for uses in VHTR and the similar systems. (auth.)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

Neutronics of a liquid salt cooled - very high temperature reactor

International Nuclear Information System (INIS)

Zakova, J.

2007-01-01

During last few years, the interest in the innovative, Liquid Salt cooled - Very High Temperature Reactor (LS-VHTR), has been growing. The preconceptual design of the LS-VHTR was suggested in Oak Ridge National Laboratory (ORNL) [1] and nowadays, several research institutions contribute to the development of this concept. The LS-VHTR design utilises a prismatic, High Temperature Reactor (HTR) fuel [2] in combination with liquid salt as a coolant. This connection of high-performance fuel and a coolant with enhanced heat transfer abilities enables efficient and economical operation. Main objective of the LS-VHTR operation may be either an efficient electricity production or a heat supply for a production of hydrogen or, combination of both. The LS-VHTR is moderated by graphite. The graphite matrix of the fuel blocks, as well as the inner and outer core reflectors serve as a thermal buffer in case of an accident, and they provide a strong thermal feedback during normal reactor operation. The high inherent safety of the LS-VHTR meets the strict requirements on future reactor systems, as defined by the Gen IV project. This work, purpose, scope, contribution to the state-of-art: The design, used in the present work is based on the first ORNL suggestion [1]. Recent study is focused on comparison of the neutronic performance of two types of fuel in the LS-VHTR core, whereas, in all previous works, only uranium fuel has been investigated. The first type of fuel, which has been employed in the present analysis, is based on the spent Light Water Reactor (LWR) fuel, whereas the second one consists of enriched uranium oxide. The results of such a comparison bring a valuable knowledge about limits and possibilities of the LS-VHTR concept, when employed as a spent fuel burner. Method:It is used a 3-D drawing of the LS-VHTR core, which contains 324x10 hexagonal fuel blocks. Each fuel block contains 216x10 fuel pins, which consists of TRISO particles incorporated into a graphite

Development of very high temperature reactor design technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, Jan Man

2012-04-01

or an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, fission product/tritium transport analysis, core thermo-fluid analysis, system layout analysis, graphite structure seismic analysis and hydrogen exposion analysis, and they are being verified and validated through a lot of international collaborations

Development of Very High Temperature Reactor Design Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, K. S.

2007-05-01

To develop design technologies for the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature for an efficient hydrogen production, key studies were performed, which include evaluation technology for the performance and safety of VHTR, and development of design and analysis codes. First, to evaluate the performance of VHTR, a series of analyses has been carried out for core characteristics at 950 .deg. C, cooled-vessel adopting internal flow path through the graphite structure, compact heat exchanger with periodic channel configuration, intermediate loop system, risk/performance-informed method, and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC, safety evaluation of both VHTR and RCCS has been also performed. In addition, prototype codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

Experimental study of core bypass flow in a prismatic VHTR based on a two-layer block model

Energy Technology Data Exchange (ETDEWEB)

Wang, Huhu, E-mail: huhuwang@tamu.edu; Hassan, Yassin A., E-mail: y-hassan@tamu.edu; Dominguez-Ontiveros, Elvis, E-mail: elvisdom@tamu.edu

2016-09-15

Bypass flow in a prismatic very high temperature gas-cooled nuclear reactor (VHTR) plays an important role in determining the coolant distribution in the core region. Efficient removal of heat from the core relies on the majority of coolant passing through the coolant channels instead of the bypass gaps. Consequently, the bypass flow fraction and its flow characteristic are important in the design process of the prismatic VHTR. The objective of this study is to experimentally investigate the flow behavior including the turbulence characteristics inside the bypass gaps using laser Doppler velocimetry (LDV), bypass fraction and pressure drops in the system. The experiment facility constructed at Texas A&M University is a scaled model consisting of two layers of fuel blocks. The distributions of the mean streamwise velocity, turbulence intensity and turbulence kinetic energy within the bypass gap at two different elevations under different Reynolds number were investigated. Uncertainties in the bypass flow fraction estimation were evaluated. The velocity and turbulence study in this work is considered to be unique, and may serve as a benchmark for the related numerical calculations.

A study on the aseismic safety of the experimental VHTR on the dense sandy layer

International Nuclear Information System (INIS)

Fujita, Shigeki; Ito, Yoshio; Baba, Osamu; Suzuki, Hideyuki; Takewaki, Naonobu; Kondo, Tsukasa; Yoshimura, Takashi; Yamada, Hitoshi.

1986-12-01

A series of studies has been carried out in 1983 and 1985 for the purpose of confirming the aseismic safety of the Experimental VHTR on the dense sandy layer. In 1983, effect of some of soil properties on seismic responses of the reactor building was estimated by means of parametric survey, and soil properties were estimated by analyzing the obserbed earthquake record. In 1985, literature review, linear, nonlinear parametric analyses and nonlinear simulation analyses were carried to study and compare the analysis method. In addition, seismic response of proposed construction site was estimated with nonlinear analysis method. As a result of these studies, the seismic response of reactor building on the dense sandy layers and wave propagation characteristics of sandy layers are understood. Especially, by means of many parametric studies, the effect of input wave characteristics, soil stiffness, nonlinear characteristics of soil properties and nonlinear analysis method on the reactor building responses were evaluated. (author)

Pebble bed modular reactor - The first Generation IV reactor to be constructed

International Nuclear Information System (INIS)

Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

2004-01-01

Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

Next Generation Nuclear Plant Materials Research and Development Program Plan

International Nuclear Information System (INIS)

G.O. Hayner; R.L. Bratton; R.N. Wright

2005-01-01

The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R and D) Program is responsible for performing R and D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R and D Program includes the following elements: (1) Developing a specific approach, program plan and other project management

Back up core designs for the experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

1979-02-01

For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

Liquid Salt Heat Exchanger Technology for VHTR Based Applications

Energy Technology Data Exchange (ETDEWEB)

Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

2012-10-11

The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2001-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

Energy Technology Data Exchange (ETDEWEB)

D. M. McEligot; K.G. Condie; G. E. Mc Creery; H. M. Mc Ilroy

2005-09-01

The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.

The shifting study of the active core or a VHTR based on the TRISO packing fraction changing

Energy Technology Data Exchange (ETDEWEB)

Silva, F.C.; Pereira, C.; Veloso, M.A.F., E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear. Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Costa, A.L. [Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (CNPq), Brasilia, DF (Brazil)

2011-07-01

A simplified VHTR core was analyzed, loaded with a fuel mixture of uranium oxide together with reprocessed transuranic nuclides. The TRUs were reprocessed together with Pu, Am, Np and Cm (23.80%) from PWR spent fuel, and dissolved in depleted uranium (0.2% {sup 235}U) until obtain 15% LEU-fuel ({sup 235}U, {sup 239}Pu and {sup 241}Pu). The shifting study of the active core was based on changes in the TRISO particle. Five cases were analyzed changing the VM/VF ratio (moderator volume/ fuel volume) making changes in the TRISO packing fraction (tpf), where tpf represents the ratio of TRISO particle on the fuel pin. The fuels were evaluated during the burnup up to 100,000.0 MWd/THM, during 990 days and without reloads. Then, it evaluated the multiplication (k{sub eff}) at zero and full power, fuel temperature coefficient ({alpha}{sub TF}), moderator temperature coefficient ({alpha}{sub TM}), and fuel composition at BOL (begin of life) and EOL (end of life), using the code Winfrith Improved Multi-Group Scheme (WIMSD5). The results show an overall heavy metal decrease in relation to the total TRU, with some Pu and Np being transmuted in the VHTR core. The results also clearly show the advantage of using reprocessed fuel in VHTR. It decreases the impact of the final spent fuel deposition, minimizes the cost of new fuel using reprocessed fuel and depleted uranium and demonstrated the promising neutronic behavior of the new types of nuclear reactors. (author)

NGNP Data Management and Analysis System Modeling Capabilities

Energy Technology Data Exchange (ETDEWEB)

Cynthia D. Gentillon

2009-09-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and identifying relationships among the measured quantities that contribute to their understanding.

NGNP Data Management and Analysis System Modeling Capabilities

International Nuclear Information System (INIS)

Gentillon, Cynthia D.

2009-01-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and identifying relationships among the measured quantities that contribute to their understanding.

Fission reactors and materials

International Nuclear Information System (INIS)

Frost, B.R.T.

1981-12-01

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

CFD analysis of the VHTR prismatic core with variation of geometry parameters

Energy Technology Data Exchange (ETDEWEB)

Lira, Carlos A.B.O.; Paiva, Pedro P.D.S., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-11-01

The Very High Temperature Reactor is a thermal, graphite moderated and helium cooled nuclear reactor. The purpose of this work is to study the behavior of the VHTR by means of parametric analysis, altering the energy generation profile in the fuel blocks and the influence of modifications in the geometry itself. The coolant flow through the coolant channels and by-pass channels were analyzed in a 1/12{sup th} section of a fuel block column. Geometry was used with by-pass channels of different dimensions, besides one that had only the cooling channels, without by-pass channel. It has been found that the existence of a by-pass flow induces an increase in the temperature gradient in the fuel block. Comparative studies were performed between the results obtained in simulations carried out with different profiles of thermal energy generation (uniform and sinusoidal) in the fuel channels. It was verified that when there is the same total thermal energy generation in the fuel block, the maximum temperature observed in each of the materials is smaller for the generation with sinusoidal profile. Computer simulations were performed using a geometry with a central channel with the same diameter as the others to verify the hypothesis that the existence of a temperature gradient in the fuel block, with the highest temperature at the center and the lowest temperature being at the periphery of this block, is due to the smaller dimension of the coolant channel located in the center of this block. The results obtained confirm the hypothesis. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-11-15

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

Programmes and projects for high-temperature reactor development

International Nuclear Information System (INIS)

Bogusch, Edgar; Hittner, Dominique

2009-01-01

An increasing attention has to be recognised worldwide on the development of High-Temperature Reactors (HTR) which has started in Germany and other countries in the 1970ies. While pebble bed reactors with spherical fuel elements have been developed and constructed in Germany, countries such as France, the US and Russia investigated HTR concepts with prismatic block-type fuel elements. The concept of a modular HTR formerly developed by Areva NP was an essential basis for the HTR-10 in China. A pebble bed HTR for electricity production is developed in South Africa. The construction is planned after the completion of the licensing procedure. Also the US is planning an HTR under the NGNP (Next Generation Nuclear Plant) Project. Due to the high temperature level of the helium coolant, the HTR can be used not only for electricity production but also for supply of process heat. Including its inherent safety features the HTR is an attractive candidate for heat supply to various types of plants e.g. for hydrogen production or coal liquefactions. The conceptual design of an HTR with prismatic fuel elements for the cogeneration of electricity and process heat has been developed by Areva NP. On the European scale the HTR development is promoted by the RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation) project. RAPHAEL is an Integrated Project of the Euratom 6th Framework Programme for the development of technologies towards a Very High-Temperature Reactor (VHTR) for the production of electricity and heat. It is financed jointly by the European Commission and the partners of the HTR Technology Network (HTR-TN) and coordinated by Areva NP. The RAPHAEL project not only promotes HTR development but also the cooperation with other European projects such as the material programme EXTREMAT. Furthermore HTR technology is investigated in the frame of Generation IV International Forum (GIF). The development of a VHTR with helium temperatures above 900 C for the

The Effect of Hold Time on Creep-Fatigue in 9Cr-1Mo

Energy Technology Data Exchange (ETDEWEB)

Oh, Tae Young; Kim, Dae Whan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Baek, Kyoung Ho [Chungnam National University, Daejeon (Korea, Republic of)

2009-05-15

9Cr-1Mo steel is a candidate material for reactor vessel for VHTR. Because 9Cr-1Mo steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. The reactor vessel of VHTR is operated at about 450 .deg. C. At this temperature, fatigue occurs during start-up and cool-down, and creep occurs during normal operation. Creep-fatigue damage by the interaction between fatigue and creep is an important factor that limits VHTR reactor vessel life. In this study, Effect of hold time on low cycle fatigue behavior of 9Cr-1Mo at 600 .deg. C was investigated in air.

The Effect of Hold Time on Creep-Fatigue in 9Cr-1Mo

International Nuclear Information System (INIS)

Oh, Tae Young; Kim, Dae Whan; Kim, Yong Wan; Baek, Kyoung Ho

2009-01-01

9Cr-1Mo steel is a candidate material for reactor vessel for VHTR. Because 9Cr-1Mo steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. The reactor vessel of VHTR is operated at about 450 .deg. C. At this temperature, fatigue occurs during start-up and cool-down, and creep occurs during normal operation. Creep-fatigue damage by the interaction between fatigue and creep is an important factor that limits VHTR reactor vessel life. In this study, Effect of hold time on low cycle fatigue behavior of 9Cr-1Mo at 600 .deg. C was investigated in air

Very-high-temperature gas reactor environmental impacts assessment

International Nuclear Information System (INIS)

Baumann, C.D.; Barton, C.J.; Compere, E.L.; Row, T.H.

1977-08-01

The operation of a Very High Temperature Reactor (VHTR), a slightly modified General Atomic type High Temperature Gas-Cooled Reactor (HTGR) with 1600 F primary coolant, as a source of process heat for the 1400 0 F steam-methanation reformer step in a hydrogen producing plant (via hydrogasification of coal liquids) was examined. It was found that: (a) from the viewpoint of product contamination by fission and activation products, an Intermediate Heat Exchanger (IHX) is probably not necessary; and (b) long term steam corrosion of the core support posts may require increasing their diameter (a relatively minor design adjustment). However, the hydrogen contaminant in the primary coolant which permeates the reformer may reduce steam corrosion but may produce other problems which have not as yet been resolved. An IHX in parallel with both the reformer and steam generator would solve these problems, but probably at greater cost than that of increasing the size of the core support posts. It is recommended that this corrosion problem be examined in more detail, especially by investigating the performance of current fossil fuel heated reformers in industry. Detailed safety analysis of the VHTR would be required to establish definitely whether the IHX can be eliminated. Water and hydrogen ingress into the reactor system are potential problems which can be alleviated by an IHX. These problems will require analysis, research and development within the program required for development of the VHTR

Reactor pressure boundary materials

International Nuclear Information System (INIS)

Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

2002-04-01

With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

Preliminary Overview of a Helium Cooling System for the Secondary Helium Loop in VHTR-based SI Hydrogen Production Facilities

Energy Technology Data Exchange (ETDEWEB)

Shin, Youngjoon; Cho, Mintaek; Kim, Dahee; Lee, Taehoon; Lee, Kiyoung; Kim, Yongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Nuclear hydrogen production facilities consist of a very high temperature gas-cooled nuclear reactor (VHTR) system, intermediate heat exchanger (IHX) system, and a sulfur-iodine (SI) thermochemical process. This study focuses on the coupling system between the IHX system and SI thermochemical process. To prevent the propagation of the thermal disturbance owing to the abnormal operation of the SI process components from the IHX system to the VHTR system, a helium cooling system for the secondary helium of the IHX is required. In this paper, the helium cooling system has been studied. The temperature fluctuation of the secondary helium owing to the abnormal operation of the SI process was then calculated based on the proposed coupling system model. Finally, the preliminary conceptual design of the helium cooling system with a steam generator and forced-draft air-cooled heat exchanger to mitigate the thermal disturbance has been carried out. A conceptual flow diagram of a helium cooling system between the IHX and SI thermochemical processes in VHTR-based SI hydrogen production facilities has been proposed. A helium cooling system for the secondary helium of the IHX in this flow diagram prevents the propagation of the thermal disturbance from the IHX system to the VHTR system, owing to the abnormal operation of the SI process components. As a result of a dynamic simulation to anticipate the fluctuations of the secondary helium temperature owing to the abnormal operation of the SI process components with a hydrogen production rate of 60 mol·H{sub 2}/s, it is recommended that the maximum helium cooling capacity to recover the normal operation temperature of 450 .deg. C is 31,933.4 kJ/s. To satisfy this helium cooling capacity, a U-type steam generator, which has a heat transfer area of 12 m{sup 2}, and a forced-draft air-cooled condenser, which has a heat transfer area of 12,388.67 m{sup 2}, are required for the secondary helium cooling system.

One stacked-column vibration test and analysis for VHTR core

International Nuclear Information System (INIS)

Ikushima, Takeshi; Ishizuka, Hiroshi; Ide, Akira; Hayakawa, Hitoshi; Shingai, Kazuteru.

1978-07-01

This paper describes experimental results of the vibration test on a single stacked-column and compares them with the analytical results. A 1/2 scale model of the core element of a very high temperature gas-cooled reactor (VHTR) was set on a shaking table. Sinusoidal waves, response time history waves, beat wave and step wave of input acceleration 100 - 900 gal in the frequency of 0.5 to 15 Hz were used to vibrate the table horizontally. Results are as follows: (1) The column has a non-linear resonance and exhibits a hysteresis response with jump points. (2) The column vibration characteristics is similar to that of the finite beams connected with non-linear soft spring. (3) The column resonance frequency decreases with increasing input acceleration. (4) The impact force increases with increasing input acceleration and boundary gap width. (5) Good correlation in vibration behavior of the stacked-column and impact force on the boundary between test and analysis was obtained. (auth.)

Nuclear reactors: physics and materials

Energy Technology Data Exchange (ETDEWEB)

Yadigaroglu, G

2005-07-01

In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

Basic data for surveillance test on core support graphite structures for the high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Iyoku, Tatsuo; Fujimoto, Nozomu; Ishihara, Masahiro; Sawa, Kazuhiro

2007-02-01

Both of the visual inspection by a TV camera and the measurement of material properties by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structures of the HTTR but also for design of advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the initial material properties of surveillance specimens before installation and installed position of surveillance specimens in the HTTR. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Burn, G.L.; Knee', S.S.; Dowker, C.L.

1994-02-01

This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

Advanced multi-physics simulation capability for very high temperature reactors

International Nuclear Information System (INIS)

Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

2012-01-01

The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

Key technology for (V)HTR: laser beam joining of SiC

International Nuclear Information System (INIS)

Knorr, J.; Lippmann, W.; Reinecke, A.M.; Wolf, R.; Rasper, R.; Kerber, A.; Wolter, A.

2005-01-01

Laser beam joining has numerous advantages over other methods presently known. After having been developed successful for brazing silicon carbide for high temperature applications, this technology is now also available for silicon nitride. Thus the field of application of SiC and Si 3 N 4 which are very interesting materials for the nuclear sector is considerably extended thanks to this new technology. Ceramic encapsulation of fuel and absorber increases the margins for operation at very high temperatures. Additionally, without ceramic encapsulation of the main core components, it will be difficult to continue claiming non-catastrophic behaviour for the (V)HTR. (orig.)

Graphite materials for nuclear reactors

International Nuclear Information System (INIS)

Oku, Tatsuo

1991-01-01

Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

International Nuclear Information System (INIS)

Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

1978-01-01

The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

International Nuclear Information System (INIS)

Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe; Perez, Marc; Cellier, Francois; Vitali, Marie-Pierre

2006-01-01

In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO 2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO 2 kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-08-15

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Technical and management challenges associated with structural materials degradation in nuclear reactors in the future

International Nuclear Information System (INIS)

Ford, F.P.

2007-01-01

issue is compounded by the fact that some of the future GEN IV reactor designs involve fast neutron spectra, and all involve increases in temperature to the range 500 o C - 1250 o C. Comparatively little is known of the effect of, for instance, creep-fatigue interactions in high irradiation fluxes on the structural integrity of the potential materials of construction. In spite of these technical concerns there is the business management expectation that all of these reactors will experience very few materials degradation problems that might affect the economics of operation. The paper starts with a review of our present capability to predict the materials degradation modes encountered in the current BWR and PWR reactor designs. This capability is the basis for any analysis of the future degradation problems (and their mitigation) in the current reactors and in the evolutionary water-cooled reactor designs. This section concludes with an overview of assessments of future materials degradation issues that might be expected in these water-cooled reactors. These preliminary discussions are then broadened to cover some of the more obvious technical problems likely to be encountered with the more advanced GEN IV designs, such as the Very High Temperature Reactor (VHTR) and the Super Critical Water Cooled Reactor (SCWR). The article concludes with a brief discussion of some of the challenges facing the technical management/leadership, with some suggestions on how to overcome them. These challenges may become especially severe given the fact that the technical problems must be overcome in a time frame that is short compared with that taken to resolve the issues that have faced us over the last 30 years. Some specific management challenges include: The decrease in the number of experienced experimentalists and analysts over the last 10 years; The decrease in 'institutional' memory as it relates to the operation of the current reactors, and the design and construction of

Materials for nuclear reactors

International Nuclear Information System (INIS)

Banerjee, S.; Kamath, H.S.

2005-01-01

The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

International Nuclear Information System (INIS)

Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

2012-01-01

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

Materials, Turbomachinery and Heat Exchangers for Supercritical CO2 Systems

Energy Technology Data Exchange (ETDEWEB)

Anderson, Mark; Nellis, Greg; Corradini, Michael

2012-10-19

The objective of this project is to produce the necessary data to evaluate the performance of the supercritical carbon dioxide cycle. The activities include a study of materials compatibility of various alloys at high temperatures, the heat transfer and pressure drop in compact heat exchanger units, and turbomachinery issues, primarily leakage rates through dynamic seals. This experimental work will serve as a test bed for model development and design calculations, and will help define further tests necessary to develop high-efficiency power conversion cycles for use on a variety of reactor designs, including the sodium fast reactor (SFR) and very high-temperature gas reactor (VHTR). The research will be broken into three separate tasks. The first task deals with the analysis of materials related to the high-temperature S-CO{sub 2} Brayton cycle. The most taxing materials issues with regard to the cycle are associated with the high temperatures in the reactor side heat exchanger and in the high-temperature turbine. The system could experience pressures as high as 20MPa and temperatures as high as 650°C. The second task deals with optimization of the heat exchangers required by the S-CO{sub 2} cycle; the S-CO{sub 2} flow passages in these heat exchangers are required whether the cycle is coupled with a VHTR or an SFR. At least three heat exchangers will be required: the pre-cooler before compression, the recuperator, and the heat exchanger that interfaces with the reactor coolant. Each of these heat exchangers is unique and must be optimized separately. The most challenging heat exchanger is likely the pre-cooler, as there is only about a 40°C temperature change but it operates close to the CO{sub 2} critical point, therefore inducing substantial changes in properties. The proposed research will focus on this most challenging component. The third task examines seal leakage through various dynamic seal designs under the conditions expected in the S-CO{sub 2} cycle

Materials for fusion reactors

International Nuclear Information System (INIS)

Ehrlich, K.; Kaletta, D.

1978-03-01

The following report describes five papers which were given during the IMF seminar series summer 1977. The purpose of this series was to discuss especially the irradiation behaviour of materials intended for the first wall of future fusion reactors. The first paper deals with the basic understanding of plasma physics relating to the fusion reactor and presents the current state of art of fusion technology. The next two talks discuss the metals intended for the first wall and structural components of a fusion reactor. Since 14 MeV neutrons play an important part in the process of irradiation damage their role is discussed in detail. The question which machines are presently available to simulate irradiation damage under conditions similar to the ones found in a fusion reactor are investigated in the fourth talk which also presents the limitations of the different methods of simulation. In this context also discussed is the importance future intensive neutron sources and materials test reactors will have for this problem area. The closing paper has as a theme the review of the present status of research of metallic and non-metallic materials in view of the quite different requirements for different fusion systems; a closing topic is the world supply on rare materials required for fusion reactors. (orig) [de

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2002-04-01

The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2002-01-01

The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

Materials for generation-IV nuclear reactors

International Nuclear Information System (INIS)

Alvarez, M. G.

2009-01-01

Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

A System Structure for a VHTR-SI Process Dynamic Simulation Code

International Nuclear Information System (INIS)

Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

2008-01-01

The VHTR-SI process dynamic simulation code embedded in a mathematical solution engine is an application software system that simulates the dynamic behavior of the VHTR-SI process. Also, the software system supports a user friendly graphical user interface (GUI) for user input/out. Structured analysis techniques were developed in the late 1970s by Yourdon, DeMarco, Gane and Sarson for applying a systematic approach to a systems analysis. It included the use of data flow diagrams and data modeling and fostered the use of an implementation-independent graphical notation for a documentation. In this paper, we present a system structure for a VHRT-SI process dynamic simulation code by using the methodologies of structured analysis

Development of the Log-in Process and the Operation Process for the VHTR-SI Process Dynamic Simulation Code

International Nuclear Information System (INIS)

Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

2009-01-01

The VHTR-SI process is a hydrogen production technique by using Sulfur and Iodine. The SI process for a hydrogen production uses a high temperature (about 950 .deg. C) of the He gas which is a cooling material for an energy sources. The Korea Atomic Energy Research Institute Dynamic Simulation Code (KAERI DySCo) is an integration application software that simulates the dynamic behavior of the VHTR-SI process. A dynamic modeling is used to express and model the behavior of the software system over time. The dynamic modeling deals with the control flow of system, the interaction of objects and the order of actions in view of a time and transition by using a sequence diagram and a state transition diagram. In this paper, we present an user log-in process and an operation process for the KAERI DySCo by using a sequence diagram and a state transition diagram

A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

International Nuclear Information System (INIS)

Kim, Hong Chul

2005-02-01

In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

Extension of the supercritical carbon dioxide Brayton cycle for application to the Very High Temperature Reactor

International Nuclear Information System (INIS)

Moisseytsev, A.; Sienicki, J. J.

2010-01-01

An investigation has been carried out of the feasibility of applying the supercritical carbon dioxide (S-CO 2 ) Brayton cycle to the Very High Temperature Reactor (VHTR). Direct application of the standard S-CO 2 recompression cycle to the VHTR was found to be challenging because of the mismatch in the inherent temperature drops across the He and CO 2 sides of the reactor heat exchanger resulting in a relatively low cycle efficiency of 45 % compared to 48 % for a direct helium cycle. Two approaches consisting of either a cascaded cycle arrangement with three separate cascaded S-CO 2 cycles or, alternately, operation of a single S-CO 2 cycle with the minimum pressure below the critical pressure and the minimum temperature above the critical temperature have been identified and shown to successfully enable the S-CO 2 Brayton cycle to be adapted to the VHTR such that the benefits of the higher S-CO 2 cycle efficiency can be realized. For both approaches, S-CO 2 cycle efficiencies in excess of 49 % are calculated. (authors)

Corrosion of structural materials for Generation IV systems

International Nuclear Information System (INIS)

Balbaud-Celerier, F.; Cabet, C.; Courouau, J.L.; Martinelli, L.; Arnoux, P.

2009-01-01

The Generation IV International Forum aims at developing future generation nuclear energy systems. Six systems have been selected for further consideration: sodium-cooled fast reactor (SFR), gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR) and very high temperature reactor (VHTR). CEA, in the frame of a national program, of EC projects and of the GIF, contributes to the structural materials developments and research programs. Particularly, corrosion studies are being performed in the complex environments of the GEN IV systems. As a matter of fact, structural materials encounter very severe conditions regarding corrosion concerns: high temperatures and possibly aggressive chemical environments. Therefore, the multiple environments considered require also a large diversity of materials. On the other hand, the similar levels of working temperatures as well as neutron spectrum imply also similar families of materials for the various systems. In this paper, status of the research performed in CEA on the corrosion behavior of the structural material in the different environments is presented. The materials studied are either metallic materials as austenitic (or Y, La, Ce doped) and ferrito-martensitic steels, Ni base alloys, ODS steels, or ceramics and composites. In all the environments studied, the scientific approach is identical, the objective being in all cases the understanding of the corrosion processes to establish recommendations on the chemistry control of the coolant and to predict the long term behavior of the materials by the development of corrosion models. (author)

Pressurized water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

Development of technologies for nuclear reactors of small and medium sized

International Nuclear Information System (INIS)

2011-08-01

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su-Jong, E-mail: paper80@snu.ac.r [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of); Lee, Sang-Moon [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Tak, Nam-il; Kim, Min-Hwan [Korea Atomic Energy Research Institute, 150-1 Deokjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, San 56-1, Daehak-Dong, Kwanak-Gu, Seoul 151-742 (Korea, Republic of)

2013-09-15

Highlights: • Bypass flow reduction method was developed by applying labyrinth seal mechanism. • Grooves on side walls of replaceable reflector block were made. • Design of the grooved wall of the reflector block was optimized by the RSA method. • The flow resistance of the bypass gap rose from 18.04 to 26.24 by the optimization. • The bypass ratios at the inlet and outlet were reduced by 36.19% and 14.66%, respectively. -- Abstract: Core bypass flow in block type very high temperature reactor (VHTR) occurs due to the inevitable gaps between the hexagonal core blocks for the block installation and refueling. Since the core bypass flow affects the reactor safety and efficiency, it should be minimized to enhance the core thermal margin. In this regard, the core bypass flow reduction method applying the labyrinth seal mechanism was developed and optimized by using the single-objective shape optimization method. Response surface approximation (RSA) method was adopted as the optimization method. Side wall of the replaceable reflector block was redesigned and response surface approximate model was adopted to optimize the shape of the reflector wall. Computational fluid dynamics (CFD) analyses were carried out not only to assess the limitation of existing method of bypass flow reduction, but also to optimize the design of a newly developed reduction method. The experiment with Seoul National University (SNU) multi-block experimental facility was performed to demonstrate the performance of the reduction method. It was found that the effect of the existing bypass flow reduction method by sealing the bypass gap exit was restricted nearby the lower region of the core. However, the flow resistance factor of the bypass gap increased from 18.04 to 26.24 by the optimized reduction method. The results of the performance test showed that the bypass flow distribution was reduced throughout the entire core regions. The bypass flow ratios at the inlet and the outlet were

Fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

none,

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

Fusion reactor materials

International Nuclear Information System (INIS)

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

Boiling water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

Advanced Reactors Around the World

International Nuclear Information System (INIS)

Majumdar, Debu

2003-01-01

At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India

Simulator for materials testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

2013-06-01

A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Polymer materials for fusion reactors

International Nuclear Information System (INIS)

Yamaoka, H.

1993-01-01

The radiation-resistant polymer materials have recently drawn much attention from the viewpoint of components for fusion reactors. These are mainly applied to electrical insulators, thermal insulators and structural supports of superconducting magnets in fusion reactors. The polymer materials used for these purposes are required to withstand the synergetic effects of high mechanical loads, cryogenic temperatures and intense nuclear radiation. The objective of this review is to summarize the anticipated performance of candidate materials including polymer composites for fusion magnets. The cryogenic properties and the radiation effects of polymer materials are separately reviewed, because there is only limited investigation on the above-mentioned synergetic effects. Additional information on advanced polymer materials for fusion reactors is also introduced with emphasis on recent developments. (orig.)

A probabilistic consequence assessment for a very high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

2017-02-15

Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

Lower activation materials and magnetic fusion reactors

International Nuclear Information System (INIS)

Conn, R.W.; Bloom, E.E.; Davis, J.W.; Gold, R.E.; Little, R.; Schultz, K.R.; Smith, D.L.; Wiffen, F.W.

1984-01-01

Radioactivity in fusion reactors can be effectively controlled by materials selection. The detailed relationship between the use of a material for construction of a magnetic fusion reactor and the material's characteristics important to waste disposal, safety, and system maintainability has been studied. The quantitative levels of radioactivation are presented for many materials and alloys, including the role of impurities, and for various design alternatives. A major outcome has been the development of quantitative definitions to characterize materials based on their radioactivation properties. Another key result is a four-level classification scheme to categorize fusion reactors based on quantitative criteria for waste management, system maintenance, and safety. A recommended minimum goal for fusion reactor development is a reference reactor that (a) meets the requirements for Class C shallow land burial of waste materials, (b) permits limited hands-on maintenance outside the magnet's shield within 2 days of a shutdown, and (c) meets all requirements for engineered safety. The achievement of a fusion reactor with at least the characteristics of the reference reactor is a realistic goal. Therefore, in making design choices or in developing particular materials or alloys for fusion reactor applications, consideration must be given to both the activation characteristics of a material and its engineering practicality for a given application

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

2016-09-26

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

WWER-1000 reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA

Corrosion of reactor materials

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-01-15

Much operational experience and many experimental results have accumulated in recent years regarding corrosion of reactor materials, particularly since the 1958 Geneva Conference on the Peaceful Uses of Atomic Energy, where these problems were also discussed. It was, felt that a survey and critical appraisal of the results obtained during this period had become necessary and, in response to this need, IAEA organized a Conference on the Corrosion of Reactor Materials at Salzburg, Austria (4-9 June 1962). It covered many of the theoretical, experimental and engineering problems relating to the corrosion phenomena which occur in nuclear reactors as well as in the adjacent circuits

Overview of gas cooled reactors' applications with CATHARE

International Nuclear Information System (INIS)

Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

2005-01-01

Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

Status of advanced nuclear reactor development in Korea

International Nuclear Information System (INIS)

Kim, H.R.; Kim, K.K.; Kim, Y.W.; Joo, H.K.

2014-01-01

The Korean nuclear industry is facing new challenges to solve the spent fuel storage problem and meet the needs to diversify the application areas of nuclear energy. In order to provide solutions to these challenges, the Korea Atomic Energy Research Institute (KAERI) has been developing advanced nuclear reactors including a Sodium-cooled Fast Reactor, Very High Temperature Gas cooled Reactor (VHTR), and System-integrated Modular Advanced Reactor (SMART) with substantially improved safety, economics, and environment-friendly features. A fast reactor system is one of the most promising options for a reduction of radioactive wastes. The long-term plan for Advanced SFR development in conjunction with the pyro-process was authorized by the Korean Atomic Energy Commission in 2008. The development milestone includes specific design approval of a prototype SFR by 2020, and the construction of a prototype SFR by 2028. KAERI has been carrying out the preliminary design of a 150MWe SFR prototype plant system since 2012. The development of advanced SFR technologies and the basic key technologies necessary for the prototype SFR are also being carried out. By virtue of high-temperature heat, a VHTR has diverse applications including hydrogen production. KAERI launched a nuclear hydrogen project using a VHTR in 2006, which focused on four basic technologies: the development of design tools, very high-temperature experimental technology, TRISO fuel fabrication, and Sulfur-iodine thermo-chemical hydrogen production technology. The technology development project will be continued until 2017. A conceptual reactor design study was started in 2012 as collaboration between industry and government to enhance the early-launching of the nuclear hydrogen development and demonstration (NHDD) project. The goal of the NHDD project is to design and build a nuclear hydrogen demonstration system by 2030. KAERI has developed SMART which is a small-sized advanced integral reactor with a rated

Influence of heating rate on corrosion behavior of Ni-base heat resistant alloys in simulated VHTR helium environment

International Nuclear Information System (INIS)

Kurata, Yuji; Kondo, Tatsuo

1985-04-01

The influence of heating rate on corrosion and carbon transfer was studied for Ni-base heat resistant alloys exposed to simulated VHTR(very high temperature reactor) coolant environment. Special attention was focused to relationship between oxidation and carburization at early stage of exposure. Tests were conducted on two heats of Hastelloy XR with different boron(B) content and the developmental alloys, 113MA and KSN. Two kinds of heating rates, i.e. 80 0 C/min and 2 0 C/min, were employed. Corrosion tests were carried out at 900 0 C up to 500 h in JAERI Type B helium, one of the simulated VHTR primary coolant specifications. Under higher heating rate, oxidation resistance of both heats of Hastelloy XR(2.8 ppmB and 40 ppmB) were equivalent and among the best, then KSN and 113MA followed in the order. Under lower heating rate only alloy, i.e. Hastelloy XR with 2.8 ppmB, showed some deteriorated oxidation resistance while all others being unaffected by the heating rate. On the other hand the carbon transfer behavior showed strong dependence on the heating rate. In case of higher heating rate, significant carburization occured at early stage of exposure and thereafter the progress of carburization was slow in all the alloys. On the other hand only slow carburization was the case throughout the exposure in case of lower heating rate. The carburization in VHTR helium environment was interpreted as to be affected by oxide film formation in the early stage of exposure. The carbon pick-up was largest in Hastelloy XR with 40 ppmB and it was followed by Hastelloy XR with 2.8 ppmB. 113MA and KSN were carburized only slightly. The observed difference of carbon pick-up among the alloys tested was interpreted to be attributed mainly to the difference of the carbon activity, the carbide precipitation characteristics among the alloys tested. (author)

New materials in nuclear fusion reactors

International Nuclear Information System (INIS)

Iwata, Shuichi

1988-01-01

In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

International Nuclear Information System (INIS)

Marmier, Alain

2012-01-01

-state conditions, both particle failure fractions (calculated with the CRYSTAL code) and the fissile material cost were determined: Wallpaper type of fuel impacts positively on the fuel cycle and reduces both the need for fissile material and the production of minor actinides, facilitating fuel reprocessing and reducing fuel cost. Safety is also improved with particle temperature being reduced during steady-state operation. This reduces the expected particle failure fraction by up to 85% over its in-core lifetime, and also the fission product release. In the long term, the VHTR is believed to be the most suitable concept for co-generation of process heat. However, in recent years the tendency in international projects goes back to lower reactor outlet temperatures mainly for 3 reasons: (1) two of the main driver countries of the VHTR have dropped thermochemical hydrogen production from their high priority list, due to expected economic and material corrosion issues; (2) the extremely high reactor outlet temperatures of > 950 degrees C would require totally new materials and construction standards; (3) the existing process heat market (steam < 600 degrees C) is already so big that investment into longer term concepts receives lower priority. In addition, the high temperature operation would tend to increase particle failure fraction and fission product release. In order to reach the VHTR objectives of high temperature and power conversion efficiency much earlier, a novel approach was developed: The nuclear part of the power plant would run at acceptably low temperature and would power a compression heat pump system acting as a temperature booster. Thus, very high temperature operation could be limited to a section of a conventional external gas circuit, avoiding the constraints related to the combination of very high temperatures and irradiation. (author)

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

International Nuclear Information System (INIS)

Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

2013-01-01

The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Models development for the fuel design of a reactor GT-MHR

International Nuclear Information System (INIS)

Telesforo R, D.; Francois L, J. L.

2009-10-01

The very high temperature reactor (VHTR) it as arisen as an option for the following reactors generation, due to their characteristics as they are inherent security, modularity and relative low cost. One of the VHTR variants, for its fuel based on prismatic blocks, is the modular reactor cooled by gas GT-MHR that uses a fuel particle of multiple layers called TRISO. These particles are small containers with fuel in their interior and they function as small pressure vessels that retain the fission products. They are absorbed inconstant ing in a cylindrical graphite matrix to form the fuel named Compact; the quantity of fuel inventory is proportional to the packaging fraction on the total volume of the Compact. The reactor consists of a matrix of 12 X 12 graphite hexagonal assemblies contained in a cylinder of 3.5 radio meters and 10.0 meters high. The nucleus has ten axial regions with 36 X 3 fuel assemblies distributed in three rings. For the neutronic modeling of the fuel and the nucleus it was employee the Monte Carlo method, using the code MCNPX (Monte Carlo N-Particle version X) that is a transport code of general purposes that uses this method with a great versatility in the representation of arbitrary three-dimensional configurations and materials configuration. The heterogeneous model of the reactor GT-MHR core was obtained, adjusting the relative parameters at core prototype GT-MHR presented by General Atomics. To prove the model it was employee the fuel formed by TRISO particles, with a nucleus of 150 μm and packaging fraction of 37.55%, with a Uranium-235 mixture, as fissile nuclide, and Thorium-232. To create a simplified model of the nucleus, or homogeneous model, without modeling any particle, it was employee the reactivity-equivalent physical transformation method that captures the effects of the heterogeneity double of the fuel region in two homogeneous equivalent cells, being obtained very good results. (Author)

A Quantitative Accident Sequence Analysis for a VHTR

Energy Technology Data Exchange (ETDEWEB)

Kim, Jintae; Lee, Joeun; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

In Korea, the basic design features of VHTR are currently discussed in the various design concepts. Probabilistic risk assessment (PRA) offers a logical and structured method to assess risks of a large and complex engineered system, such as a nuclear power plant. It will be introduced at an early stage in the design, and will be upgraded at various design and licensing stages as the design matures and the design details are defined. Risk insights to be developed from the PRA are viewed as essential to developing a design that is optimized in meeting safety objectives and in interpreting the applicability of the existing demands to the safety design approach of the VHTR. In this study, initiating events which may occur in VHTRs were selected through MLD method. The initiating events were then grouped into four categories for the accident sequence analysis. Initiating events frequency and safety systems failure rate were calculated by using reliability data obtained from the available sources and fault tree analysis. After quantification, uncertainty analysis was conducted. The SR and LR frequency are calculated respectively 7.52E- 10/RY and 7.91E-16/RY, which are relatively less than the core damage frequency of LWRs.

Reactor simulator development. Workshop material

International Nuclear Information System (INIS)

2001-01-01

The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIM tm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

International Nuclear Information System (INIS)

Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

1977-12-01

Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)

Chemical reactor for converting a first material into a second material

Science.gov (United States)

Kong, Peter C

2012-10-16

A chemical reactor and method for converting a first material into a second material is disclosed and wherein the chemical reactor is provided with a feed stream of a first material which is to be converted into a second material; and wherein the first material is combusted in the chemical reactor to produce a combustion flame, and a resulting gas; and an electrical arc is provided which is passed through or superimposed upon the combustion flame and the resulting gas to facilitate the production of the second material.

Developments in reactor materials science methodology

International Nuclear Information System (INIS)

Tsykanov, V.A.; Ivanov, V.B.

1987-01-01

Problems related to organization of investigations into reactor materials science are considered. Currently the efficiency and reliability of nuclear power units are largely determined by the fact, how correctly and quickly conclusions concerning the parameters of designs and materials worked out for a long time in reactor cores, are made. To increase information value of materials science investigations it is necessary to create a uniform system, providing for solving methodical, technical and organizational problems. Peculiarities of the current state of reactor material science are analysed and recommendations on constructing an optimal scheme of investigations and data flow interconnection are given

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

Research reactors and materials testing

International Nuclear Information System (INIS)

Vidal, H.

1986-01-01

Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Evaluation of the oxidation behavior and strength of the graphite components in the VHTR, (1)

International Nuclear Information System (INIS)

Eto, Motokuni; Kurosawa, Takeshi; Nomura, Shinzo; Imai, Hisashi

1987-04-01

Oxidation experiments have been carried out mainly on a fine-grained isotropic graphite, IG-110, at temperatures between 1173 and 1473 K in a water vapor/helium mixture. In most cases water vapor concentration was 0.65 vol% and helium pressure, 1 atm. Reaction rate and burn-off profile were measured using cylindrical specimens. On the basis of the experimental data the oxidation behavior of fuel block and core support post under the condition of the VHTR operation was estimated using the first-order or Langmuir-Hinshelwood equation with regard to water vapor concentration. Strength and stress-strain relationship of the graphite components with burn-off profiles estimated above were analyzed on the basis of the model for stress-strain relationship and strength of graphite specimens with density gradients. The estimation indicated that the integrity of the components would be maintained during normal reactor operation. (author)

Failure Predictions for Graphite Reflector Bricks in the Very High Temperature Reactor with the Prismatic Core Design

Energy Technology Data Exchange (ETDEWEB)

Singh, Gyanender, E-mail: sing0550@umn.edu [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Fok, Alex [Minnesota Dental Research in Biomaterials and Biomechanics, School of Dentistry, University of Minnesota, 515, Delaware St. SE, Minneapolis, MN 55455 (United States); Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Mantell, Susan [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States)

2017-06-15

Highlights: • Failure probability of VHTR reflector bricks predicted though crack modeling. • Criterion chosen for defining failure strongly affects the predictions. • Breaching of the CRC could be significantly delayed through crack arrest. • Capability to predict crack initiation and propagation demonstrated. - Abstract: Graphite is used in nuclear reactor cores as a neutron moderator, reflector and structural material. The dimensions and physical properties of graphite change when it is exposed to neutron irradiation. The non-uniform changes in the dimensions and physical properties lead to the build-up of stresses over the course of time in the core components. When the stresses reach the critical limit, i.e. the strength of the material, cracking occurs and ultimately the components fail. In this paper, an explicit crack modeling approach to predict the probability of failure of a VHTR prismatic reactor core reflector brick is presented. Firstly, a constitutive model for graphite is constructed and used to predict the stress distribution in the reflector brick under in-reactor conditions of high temperature and irradiation. Fracture simulations are performed as part of a Monte Carlo analysis to predict the probability of failure. Failure probability is determined based on two different criteria for defining failure time: A) crack initiation and B) crack extension to near control rod channel. A significant difference is found between the failure probabilities based on the two criteria. It is predicted that the reflector bricks will start cracking during the time range of 5–9 years, while breaching of the control rod channels will occur during the period of 11–16 years. The results show that, due to crack arrest, there is a significantly delay between crack initiation and breaching of the control rod channel.

A Sub-channel Analysis of a VHTR Fuel Block with Tin Gap-Filler

International Nuclear Information System (INIS)

Cho, Chung Ho; Kim, Yong Hee; Yi, Yong Sun; Kim, Hong Pyo

2005-01-01

In the Nuclear Hydrogen Development and Demonstration (NHDD) project, two types of VHTRs (Very High Temperature Reactors), prismatic or pebble bed, are under investigation as the nuclear heat source for hydrogen production. In general, the targeted coolant outlet temperature of VHTR is 950∼1000 .deg. C and the maximum allowable fuel temperature is 1250 .deg. C during the normal operation. In the case of the prismatic reactor (PMR), conventional fuel designs result in a small margin in the maximum fuel temperature. This is one of the biggest demerits of the prismatic type In this paper, a technique of lowering the maximum fuel temperature is suggested. The PMR fuel assembly is comprised of many coolant holes and fuel channels. Cylindrical fuel compacts are stacked inside the fuel channel. Consequently, there should be a physical gap between the fuel compact and graphite block, which is filled with the He gas in the conventional design. The heat transfer coefficient of the He gap is very poor, and this increases the fuel temperature substantially. In the proposed design measure, the gap is filled with a liquid metal, tin (Sn) that has a very high thermal conductivity. The effects of tin in the gap with gap distance variation in the viewpoint of thermal hydraulics are quantitatively discussed. Also, the effects of the variations of the axial power distribution are discussed

Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

International Nuclear Information System (INIS)

1976-08-01

This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

Assessment of the crossflow loss coefficient in Very High Temperature Reactor core - 15338

International Nuclear Information System (INIS)

Lee, S.N.; Tak, N.I.; Kim, M.H.; Noh, J.M.

2015-01-01

The Very High Temperature Reactor (VHTR) is a helium gas cooled and graphite moderated reactor. It was chosen as one of the Gen-4 reactors owing to its inherent safety. Various researches for prismatic gas-cooled reactors have been conducted for efficient and safe use. The prismatic VHTR consists of vertically stacked fuel blocks. Between the vertical fuel blocks, there is cross gap because of manufacturing tolerance or graphite change during the operation. This cross gap changes the coolant flow path, called a crossflow, which may affect the fuel temperature. Various tests and numerical studies have been conducted to predict the crossflow and loss coefficient. In the present study, the CFD calculation is conducted to draw the loss coefficient, and compared with Groehn, Kaburaki and General Atomics (GA) correlations. The results of the Groehn and Kaburaki correlations tend to decrease as the gap size increases, whereas the data of GA show the opposite. The loss coefficient given by the CFD calculation tends to maintain the regular value without regard to the gap size for the standard fuel block, like the Groehn correlation. However, the loss coefficient of the control fuel block increases as the gap size widens, like the GA results

Research on Evaluation Methodology for High Temperature Components and Technical Issues

International Nuclear Information System (INIS)

Kim, Y.J.; Han, S.B.

2007-03-01

The research on evaluation methodology for high temperature components and technical issues includes the comparison of evaluation technology of Very High Temperature Reactors(VHTRs) with that of present commercial reactors, the review of Hot Gas Duct(HGD) insulation designs, the analysis of the codes related to VHTR component construction and the analysis of technical issues on application of present codes to HGD construction. Codes to assure the integrity of the VHTR components are not fully prepared yet in any country. To understand the evaluation technology of the VHTR-related codes, key requirements of ASME B and PV Code Section III, Subsection NB and NH were compared. Six kinds of HGD designs were reviewed and compared. A reference which analyzed seven kinds of present component codes were reviewed and the limitations of them were summarized. Especially it was found that the selection of materials is limited, material property data are not enough, and design analysis methodology is not fully specified

Gas-cooled reactors: the importance of their development

International Nuclear Information System (INIS)

Kasten, P.R.

1979-06-01

The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

Real time simulator for material testing reactor

Energy Technology Data Exchange (ETDEWEB)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

2012-03-15

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

Real time simulator for material testing reactor

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

2012-01-01

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

Reactor core materials research and integrated material database establishment

International Nuclear Information System (INIS)

Ryu, Woo Seog; Jang, J. S.; Kim, D. W.

2002-03-01

Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

A study on bypass flow gap distribution in a prismatic VHTR core

International Nuclear Information System (INIS)

Kim, M. H.; Jo, C. K.; Lim, H. S.

2010-01-01

Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of irradiation fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass flow and the location of core hot spots are closely related and a measure to reduce the bypass flow is necessary. (authors)

Oxidation kinetics of innovative carbon materials with respect to severe air ingress accidents in HTRs and graphite disposal or processing

International Nuclear Information System (INIS)

Schloegel, Baerbel

2010-01-01

Currently future nuclear reactor concepts of the Fourth Generation (Gen IV) are under development. To some extend they apply with new, innovative materials developed just for this purpose. This thesis work aims at a concept of Generation IV Very High Temperature Reactors (VHTR) in the framework of the European project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation). The concept named ANTARES (AREVA New Technology based on advanced gas-cooled Reactors for Energy Supply) was developed by AEVA NP. It is a helium cooled, graphite moderated modular reactor for electricity and hydrogen production, by providing the necessary process heat due to its high working temperature. Particular attention is given here to oxidation kinetics of newly developed carbon materials (NBG-17) with still unknown but needed information in context of severe air ingress accident in VHTR's. Special interest is paid to the Boudouard reaction, the oxidation of carbon by CO 2 . In case of an air ingress accident, carbon dioxide is produced in the primary reaction of atmospheric oxygen with reflector graphite. From there CO 2 could flow into the reactor core causing further damage by conversion into CO. The purpose of this thesis is to ascertain if and to what degree this could happen. First of all oxidation kinetic data of the Boudouard reaction with NBG-17 is determined by experiments in a thermo gravimetric facility. The measurements are evaluated and converted into a common formula and a Langmuir-Hinshelwood similar oxidation kinetic equation, as input for the computer code REACT/THERMIX. This code is then applied to analyse severe air ingress accidents for several air flow rates. The results are discussed for two accident situations, in which a certain graphite burn off is achieved. All cases show much more damage to the graphite bottom reflector than to the reactor core. Thus the bottom reflector will lose its structural integrity much earlier than the core itself will

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

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

PIV Experiments to Measure Flow Phenomena in a Scaled Model of a VHTR Lower Plenum

Energy Technology Data Exchange (ETDEWEB)

Hugh M. McIlroy, Jr.; Donald M. McEligot; Richard R. Schultz; Daniel Christensen; Robert J. Pink; Ryan C. Johnson

2006-09-01

A report of experimental data collected at the Matched-Index-of-Refraction (MIR) Laboratory in support of contract DE-AC07-05ID14517 and the INL Standard Problem on measurements of flow phenomena occurring in a lower plenum of a typical prismatic VHTR concept reactor to assess CFD code is presented. Background on the experimental setup and procedures is provided along with several samples of data obtained from the 3-D PIV system and an assessment of experimental uncertainty is provided. Data collected in this study include 3-dimensional velocity-field descriptions of the flow in all four inlet jets and the entire lower plenum with inlet jet Reynolds numbers (ReJet) of approximately 4300 and 12,400. These investigations have generated over 2 terabytes of data that has been processed to describe the various velocity components in formats suitable for external release and archived on removable hard disks. The processed data from both experimental studies are available in multi-column text format.

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

International Nuclear Information System (INIS)

Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

2012-01-01

Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

International Nuclear Information System (INIS)

Ougouag, A.M.; Ferrer, R.M.

2010-01-01

The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these 'spread' the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

Materials requirements for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Bennett, J.W.; Horton, K.E.

1978-01-01

Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

Composites as structural materials in fusion reactors

International Nuclear Information System (INIS)

Megusar, J.

1989-01-01

In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

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.

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Proposal of world network on material testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

2011-01-01

Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

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

1980-10-01

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

Positron annihilation studies on structural materials for nuclear reactors

International Nuclear Information System (INIS)

Rajaraman, R.; Amarendra, G.; Sundar, C.S.

2012-01-01

Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

International Nuclear Information System (INIS)

Was, Gary S.

2007-01-01

The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

Materials design data for fusion reactors

International Nuclear Information System (INIS)

Tavassoli, A.A.F.

1998-01-01

Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.)

Materials design data for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A.A.F. [CEA Commissariat a l`Energie Atomique, Gif sur Yvette (France). CEREM

1998-10-01

Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.) 19 refs.

Next Generation Nuclear Plant Materials Selection and Qualification Program Plan

Energy Technology Data Exchange (ETDEWEB)

R. Doug Hamelin; G. O. Hayner

2004-11-01

The U.S. Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design is a graphite-moderated, helium-cooled, prismatic or pebble bed thermal neutron spectrum reactor with an average reactor outlet temperature of at least 1000 C. The NGNP will use very high burn up, lowenriched uranium, TRISO-Coated fuel in a once-through fuel cycle. The design service life of the NGNP is 60 years.

Thermochemical data for reactor materials

International Nuclear Information System (INIS)

Ronchi, C.; Turrini, F.

1990-01-01

This report describes a computer database of thermochemical properties of nuclear reactor materials to be used for source term calculations in reactor accident codes. In the first part, the structure and the content of the computer file is described. In the second part a set of thermochemical data is presented pertaining to chemical reactions occurring during severe nuclear reactor accidents and involving fuel (uranium dioxide), fission products and structural materials. These data are complementary to those collected in the databook recently published by Cordfunke and Potter after a study supported by the Commission of the European Communities. The present data were collected from review articles and databanks and follow a discussion on the uncertainties and errors involved in the calculation of complex chemical equilibria in the extrapolated temperature range

Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

2016-04-15

Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

International Nuclear Information System (INIS)

Park, HyeongKae; Knoll, Dana; Gaston, Derek; Martineau, Richard

2010-01-01

We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

International Nuclear Information System (INIS)

Allen, Todd R.; Busby, J. T.; Klueh, R. L.; Maloy, Stuart A.; Toloczko, Mychailo B.

2008-01-01

This is a review article that provides an overview of the reactor core structural materials and clad and duct needs for the GNEP advanced burner reactor design. A short history of previous research on structural materials for irradiation environments is provided. There is also a section describing some advanced materials that may be candidate materials for various reactor core structures

High temperature oxidation and corrosion behavior of Ni-base superalloy in He environment

International Nuclear Information System (INIS)

Lee, Gyoeng Geun; Park, Ji Yeon; Jung, Su jin

2010-11-01

Ni-base superalloy is considered as a IHX (Intermediate Heat Exchanger) material for VHTR (Very High Temperature Gas-Cooled Reactor). The helium environment in VHTR contains small amounts of impure gases, which cause oxidation, carburization, and decarburization. In this report, we conducted the literature survey about the high temperature behavior of Ni-base superalloys in air and He environments. The basic information of Ni-base superalloy and the basic metal-oxidation theory were briefly stated. The He effect on the corrosion of Ni-base superalloy was also summarized. This works would provide a brief suggestion for the next research topic for the application of Ni-base superalloy to VHTR

Reactor materials research as an effective instrument of nuclear reactor perfection

International Nuclear Information System (INIS)

Baryshnikov, M.

2006-01-01

The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

Compatibility of heat resistant alloys with boron carbide, 5

International Nuclear Information System (INIS)

Baba, Shinichi; Kurasawa, Toshimasa; Endow, Taichi; Someya, Hiroyuki; Tanaka, Isao.

1986-08-01

This paper includes an experimental result of out-of-pile compatibility and capsule design for irradiation test in Japan Materials Testing Reactor (JMTR). The compatibility between sheath material and neutron absorber materials for control rod devices (CRD) was examined for potential use in a very high temperature reactor (VHTR) which is under development at JAERI. The purpose of the compatibility tests are preliminary evaluation of safety prior to irradiation tests. Preliminary compatibility evaluation was concerned with three items as follows : 1) Lithium effects on the penetrating reaction of Incoloy 800H alloy in contact with a mixture of boronated graphite and lithium hydroxide powders, 2) Short term tensile properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite and fracture mode analysis, 3) Reaction behavior of both alloys under transient power conditions of a VHTR. It was clear that the reaction rate constant of the Incoloy 800H alloy was accelerated by doping lithium hydroxide into the boron carbide and graphite powder. The mechanical properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite were decreased. Ultimate tensile strength and tensile ductilities at temperatures over 850 deg C were reduced, but there was no change in the proof (yield) stress. Both alloys exhibited a brittle intergranular fracture mode during transient power conditions of a VHTR and also exhibited severe penetration. Irradiation capsules for compatibility test were designed to simulate three irradiation conditions of VHTR: 1) steady state for VHTR, 2) Transient power condition, 3) Service limited life of CRD. Capsule irradiation experiments have been carried out satisfactorily and thus confirm the validity of the capsule design procedure. (author)

Nuclear reactor structural material forming less radioactive corrosion product

International Nuclear Information System (INIS)

Nakazawa, Hiroshi.

1988-01-01

Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

2006-12-15

The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

International Nuclear Information System (INIS)

Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

2006-12-01

The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

Materials Degradation in Light Water Reactors: Life After 60,

International Nuclear Information System (INIS)

Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

2008-01-01

Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

International Nuclear Information System (INIS)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

2015-01-01

Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO 2 emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO 2 emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus TM Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition

The use of low enriched uranium fuel cycle in high temperature gas-cooled reactors

International Nuclear Information System (INIS)

The present paper begins with a brief review of the status of research and development of experimental VHTR in Japan. On the basis of the experience gained from these work, assessment is made of commercial HTRs. Material balance with fuel burnup is calculated for the two core models; one is HTGR for steam cycle and the other VHTR for process heat application. The results of assessment of commercial HTRs are compared with those for LWR

Development of C/C composite for the core component of the high temperature gas cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

2005-01-15

This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

Development of C/C composite for the core component of the high temperature gas cooled reactor

International Nuclear Information System (INIS)

Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H.

2005-01-01

This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite

Euratom research and training in generation IV systems with emphasis on V/HTR

International Nuclear Information System (INIS)

Goethem, G. van; Manolatos, P.; Fuetterer, M.

2006-01-01

In this overview paper, the following questions are addressed: (1) What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2) What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle: Education, Research, and Innovation') respond to the S/T challenges: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is principally under the responsibility of the 2 Directorates Generals (DG) DG Research (RTD, located in Brussels), which implements and manages the programme of 'indirect actions', and the DG Joint Research Centre (JRC, headquarters in Brussels and 7 scientific institutes in 5 Member States) which carries out 'direct actions' in their own laboratories. In this HTR-2006 introductory paper, the emphasis is on the indirect and direct actions of the 6 th Euratom research framework programme 2003-2006, FP-6, with special emphasis on V/HTR Generation IV research. (orig.)

Materials for high temperature reactor vessels

International Nuclear Information System (INIS)

Buenaventura Pouyfaucon, A.

2004-01-01

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

Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

Energy Technology Data Exchange (ETDEWEB)

Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

2014-09-01

High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

International Nuclear Information System (INIS)

Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

2010-01-01

The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

Overview moderator material for nuclear reactor components

International Nuclear Information System (INIS)

Mairing Manutu Pongtuluran; Hendra Prihatnadi

2009-01-01

In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

Perspectives for the french R and D program for high and very high temperature reactors - HTR2008-58172

International Nuclear Information System (INIS)

Yvon, P.; Hittner, D.; Delbecq, J. M.

2008-01-01

A R and D programme has been launched addressing the needs of the development of an indirect cycle flexible modular HTR operating at 850 deg. C for electricity generation and/or heat production for industrial processes. In the frame of this program, several significant technical challenges required to demonstrate the viability and performance of the system have been successfully addressed. Design and safety analysis needed the development of computational tools, therefore reactor physics, and thermo-fluid dynamics codes have been developed and are now in the process of being validated in the frame of international code-to-code and code to experiment benchmarks. Most importantly, the performance of the HTR/VHTR fuel identified as TRISO-coated particle must prove to be excellent in operating as well as accidental conditions. A manufacturing and quality control process has been developed and now fuel qualification based on irradiation and heating safety tests is being prepared on the basis of irradiation programs in France and in the frame of the GENERATION IV International Forum (GIF) as well as the development of fuel behaviour models including performance data, failure particle prediction and long-term integrity of the coating. Material and component technologies have been investigated in normal and accident conditions for V/HTR objectives. Significant progress has been made for vessel structures and reactor core structural elements. Major challenges still lie ahead for plate type compact intermediate heat exchangers, especially at temperatures above 850 deg. C, but an alternative solution with helical tubes is also being developed. In order to demonstrate that materials have adequate performance over long service life under impure helium environment and constraints, the research programme focuses on microstructural and mechanical property data, long-term irradiation behaviour, corrosion, modelling and codification of design rules as well as qualification of

Overview of the Modified SI Cycle to Produce Nuclear Hydrogen Coupled to VHTR

International Nuclear Information System (INIS)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

2016-01-01

The steam reforming of methane is one of hydrogen production processes that rely on cheap fossil feedstocks. An overview of the VHTR-based nuclear hydrogen production process with the modified SI cycle has been carried out to establish whether it can be adopted as a feasible technology to produce nuclear hydrogen

National Project Management Corp (United States). [Technical and regulatory aspects

International Nuclear Information System (INIS)

DeBor, Joseph

2013-01-01

Studies of the VHTR fuel cycle which involves: (i) use of light water cooled reactor (LWR) spent fuel as kernel feedstock; (ii) recycle of spent VHTR fuel; (iii) use of the VHTR in the management of transuranics (TRU); and, (iv) the geologic storage performance of spent VHTR fuel

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

Next Generation Nuclear Plant Materials Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

G. O. Hayner; E.L. Shaber

2004-09-01

The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years.

Proceedings of the international symposium on materials testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kawamura, Hiroshi

2009-01-01

This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

Hydrogen production using the sulfur-iodine cycle coupled to a VHTR: An overview

International Nuclear Information System (INIS)

Vitart, X.; Le Duigou, A.; Carles, P.

2006-01-01

The sulfur-iodine thermo-chemical cycle is considered to be one of the most promising routes for massive hydrogen production, using high temperature heat from a Generation IV VHTR. We propose here a brief overview of the main questions raised by this cycle, along with the general lines of French CEA's program

Overview of gas-cooled reactor systems, their importance and their interactions

International Nuclear Information System (INIS)

Kasten, P.R.; Spiewak, I.; Tobias, M.L.

1975-01-01

The economic interactions between fueling, separative work, and capital requirements are illustrated for HTGR, GCFR, HTGR-GT, VHTR, LWRs and LMFBRs. The influence of finite low-cost uranium resources and of extensive LWR application within the next two decades on reactor use is also discussed. Technological developments required for the practical application of HTGRs, GCFRs, HTGR-GT and VHTRs are presented, along with the importance and environmental effects features of these applications. The technical advantages and disadvantages associated with use of the uranium and the thorium fuel cycles in HTGRs are given, including the implications a given fuel cycle has on fuel recycle and mined-fuel requirements. The influence of core design on HTGR fuel and coolant temperatures and on associated performance features are illustrated by considering prismatic and pebble-bed type cores. Finally, several scenarios relative to the development of the HTGR, GCFR, HTGR-GT and VHTR are presented. (auth)

Structural materials issues for the next generation fission reactors

Science.gov (United States)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

First conceptual design of the experimental multi-purpose high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Tsunoda, T [Fuji Electric Co. Ltd., Tokyo (Japan)

1976-02-01

A part of the multi-purpose high temperature reactor (VHTR) was designed by the First Atomic Power Industry Group (FAPIG). Both Fuji Electric Co., Ltd. and Kawasaki Heavy Industries, Ltd. of the FAPIG group took charge of the design of main parts of the reactor Kobe Steel, Ltd., Ebara Manufacturing Co., Ltd., Shimizu Construction Co., Ltd. and the Nuclear Fuel Corp. have associated with this group. The reactor system includes a nuclear reactor and two cooling loops provided through intermediate heat exchangers in order to utilize the heat of helium gas delivered from the reactor outlet at 1,000 deg C. One is a reformer loop to produce the reducing gas for steel manufacture. The other is a testing loop for a reducing gas heater and a gas turbine. These loops transfer heat of about 25 MW at 930 deg C at rated capacity. The reformer can supply the reducing gas equivalent to the production of 100 tons per day sponge iron. A housing of the reactor is composed of a primary steel container, internal concrete and a secondary container made of reinforced concrete. The construction is based on the following principles. (1) For the very high temperature portion at 1,000 deg C, a non-metallic material such as graphite should be used. (2) The metallic construction shall be cooled with return gas below 400 deg C. (3) The steel pressure vessel shall be employed. (4) The design shall be based on the existing gas furnace.

Initial VHTR accident scenario classification: models and data.

Energy Technology Data Exchange (ETDEWEB)

Vilim, R. B.; Feldman, E. E.; Pointer, W. D.; Wei, T. Y. C.; Nuclear Engineering Division

2005-09-30

Nuclear systems codes are being prepared for use as computational tools for conducting performance/safety analyses of the Very High Temperature Reactor. The thermal-hydraulic codes are RELAP5/ATHENA for one-dimensional systems modeling and FLUENT and/or Star-CD for three-dimensional modeling. We describe a formal qualification framework, the development of Phenomena Identification and Ranking Tables (PIRTs), the initial filtering of the experiment databases, and a preliminary screening of these codes for use in the performance/safety analyses. In the second year of this project we focused on development of PIRTS. Two events that result in maximum fuel and vessel temperatures, the Pressurized Conduction Cooldown (PCC) event and the Depressurized Conduction Cooldown (DCC) event, were selected for PIRT generation. A third event that may result in significant thermal stresses, the Load Change event, is also selected for PIRT generation. Gas reactor design experience and engineering judgment were used to identify the important phenomena in the primary system for these events. Sensitivity calculations performed with the RELAP5 code were used as an aid to rank the phenomena in order of importance with respect to the approach of plant response to safety limits. The overall code qualification methodology was illustrated by focusing on the Reactor Cavity Cooling System (RCCS). The mixed convection mode of heat transfer and pressure drop is identified as an important phenomenon for Reactor Cavity Cooling System (RCCS) operation. Scaling studies showed that the mixed convection mode is likely to occur in the RCCS air duct during normal operation and during conduction cooldown events. The RELAP5/ATHENA code was found to not adequately treat the mixed convection regime. Readying the code will require adding models for the turbulent mixed convection regime while possibly performing new experiments for the laminar mixed convection regime. Candidate correlations for the turbulent

Experimental Investigation on Cross Flow of Wedge-shaped Gap in the core of Prismatic VHTR

International Nuclear Information System (INIS)

Lee, Jeong Hun; Park, Goon Cherl; Cho, Hyoung Kyu; Yoon, Su Jong

2014-01-01

The core of the PMR type reactor consists of assemblies of hexagonal graphite blocks. The graphite blocks have lots of advantages for neutron economy and high temperature structural integrity. The height and flat-to-flat width of fuel bock are 793 mm and 360 mm, respectively. Each block has 108 coolant channels of which the diameter is 16 mm. And there are gaps between blocks not only vertically but also horizontally for reloading of the fuel elements. The vertical gap induces the bypass flow and through the horizontal gap the cross flow is formed. Since the complicated flow distribution occurs by the bypass flow and cross flow, flow characteristics in the core of the PMR reactor cannot be treated as a simple pipe flow. The fuel zone of the PMR core consists of multiple layers of fuel blocks. The shape change of the fuel blocks could be caused by the thermal expansion and fast-neutron induced shrinkage. It could make different axial shrinkage of fuel block and this leads to wedge-shaped gaps between two stacked fuel blocks. The cross flow is often considered as a leakage flow through the horizontal gap between stacked fuel blocks and it complicates the flow distribution in the reactor core by connecting the coolant channel and the bypass gap. Moreover, the cross flow could lead to uneven coolant distribution and consequently cause superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. To develop the cross flow loss coefficient model for determination of the flow distribution for PMR core analysis codes, study on cross flow for PMR200 core is essential. In particular, to predict the amount of flow through the cross flow gap, obtaining accurate flow loss coefficient is important. In this study, the full-scale cross flow experimental facility was constructed to

Analysis of Creep Crack Growth Behavior of Alloy 617 for Use in a VHTR System

International Nuclear Information System (INIS)

Kim, Woo-Gon; Kim, Min-Hwan; Park, Jae-Young; Ekaputra, I. M. W.; Kim, Seon-Jin

2015-01-01

Alloy 617 is a major candidate material for the IHX component. The design of the component, which will operate well into the creep range, will require a good understanding of creep crack growth deformation. Efforts are now being undertaken in the Gen-IV program to provide data needed for the design and licensing of the nuclear plants, and with this goal in mind, to meet the needs of the conceptual designers of the VHTR system, 'Gen-IV Materials Handbook' is being established through an international collaboration program of GIF (Gen-IV Forum) countries. To logically obtain the B and q values in the CCGR equation, three methods in terms of LSFM, MVM, and PDM were adopted. The PDM was most useful. Both the B and q coefficients followed a lognormal distribution. Using a lognormal distribution in the PDM, a number of random variables were generated by Monte Carlo Simulation, and the CCGR lines could be successfully predicted from the viewpoint of reliability

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1977-01-01

Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

Maintenance and material aspects of DREAM reactor

International Nuclear Information System (INIS)

Ueda, S.; Nishio, S.; Yamada, R.; Seki, Y.; Kurihara, R.; Adachi, J.; Yamazaki, S.

2000-01-01

A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

Computer simulation of multi-elemental fusion reactor materials

International Nuclear Information System (INIS)

Voertler, K.

2011-01-01

Thermonuclear fusion is a sustainable energy solution, in which energy is produced using similar processes as in the sun. In this technology hydrogen isotopes are fused to gain energy and consequently to produce electricity. In a fusion reactor hydrogen isotopes are confined by magnetic fields as ionized gas, the plasma. Since the core plasma is millions of degrees hot, there are special needs for the plasma-facing materials. Moreover, in the plasma the fusion of hydrogen isotopes leads to the production of high energetic neutrons which sets demanding abilities for the structural materials of the reactor. This thesis investigates the irradiation response of materials to be used in future fusion reactors. Interactions of the plasma with the reactor wall leads to the removal of surface atoms, migration of them, and formation of co-deposited layers such as tungsten carbide. Sputtering of tungsten carbide and deuterium trapping in tungsten carbide was investigated in this thesis. As the second topic the primary interaction of the neutrons in the structural material steel was examined. As model materials for steel iron chromium and iron nickel were used. This study was performed theoretically by the means of computer simulations on the atomic level. In contrast to previous studies in the field, in which simulations were limited to pure elements, in this work more complex materials were used, i.e. they were multi-elemental including two or more atom species. The results of this thesis are in the microscale. One of the results is a catalogue of atom species, which were removed from tungsten carbide by the plasma. Another result is e.g. the atomic distributions of defects in iron chromium caused by the energetic neutrons. These microscopic results are used in data bases for multiscale modelling of fusion reactor materials, which has the aim to explain the macroscopic degradation in the materials. This thesis is therefore a relevant contribution to investigate the

Materials data base for fusion reactors-I

International Nuclear Information System (INIS)

Iwata, S.; Nogami, A.; Ishino, S.; Mishima, Y.; Takao, Y.; Aruga, T.; Shiraishi, K.

1982-01-01

The materials data base is a set of experimental and/or calculated data being compiled to meet the broad needs for materials data by taking advantage of the data base management systems. In this paper the objective of such computerized data base is described and the characteristics of fusion reactor materials are discussed from the viewpoint of the data base development. The near-term emphasis of the development has been put on the irradiation data for 316 type stainless steels. Through the test of this small data base, it can be concluded that this approach is promising for materials data base management and for the establishment of the interface between fusion reactor designer and materials investigator. (orig.)

Overview of materials research for fusion reactors

International Nuclear Information System (INIS)

Muroga, T.; Gasparotto, M.; Zinkle, S.J.

2002-01-01

Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Fusion Reactor Materials

International Nuclear Information System (INIS)

Moons, F.

1998-01-01

SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

Fuel Development For Gas-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

M. K. Meyer

2006-06-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

International Nuclear Information System (INIS)

Baumann, E.W.; Ondrejcin, R.S.

1990-11-01

Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

Critical plasma-materials issues for fusion reactor designs

International Nuclear Information System (INIS)

Wilson, K.L.; Bauer, W.

1983-01-01

Plasma-materials interactions are a dominant driving force in the design of fusion power reactors. This paper presents a summary of plasma-materials interactions research. Emphasis is placed on critical aspects related to reactor design. Particular issues to be addressed are plasma edge characterization, hydrogen recycle, impurity introduction, and coating development. Typical wall fluxes in operating magnetically confined devices are summarized. Recent calculations of tritium inventory and first wall permeation, based on laboratory measurements of hydrogen recycling, are given for various reactor operating scenarios. Impurity introduction/wall erosion mechanisms considered include sputtering, chemical erosion, and evaporation (melting). Finally, the advanced material development for in-vessel components is discussed. (author)

A Dynamic Simulation Program for a Hydriodic Acid Concentration and Decomposition Process in the VHTR-SI Process

International Nuclear Information System (INIS)

Chang, Ji Woon; Shin, Young Joon; Lee, Tae Hoon; Lee, Ki Young; Kim, Yong Wan; Chang, Jong Hwa; Youn, Cheung

2011-01-01

The Sulfur-Iodine (SI) cycle which can produce hydrogen by using nuclear heat consists of a Bunsen reaction (Section 1), a sulfur acid concentration and decomposition (Section 2), and a hydriodic acid concentration and decomposition (Section 3). The heat required in the SI process can be supplied through an intermediate heat exchanger (IHX) by a Very High Temperature Gas Cooled Reactor (VHTR). The Korea Atomic Energy Research Institute-Dynamic Simulation Code (KAERI-DySCo) based on the Visual C++ is an integration application software that simulates the dynamic behavior of the SI process. KAERI-DySCo was prepared to solve dynamic problem of the seven chemical reactors which consist of Sections 2 and 3. Section 3 is the key part of the SI process, because the strong non-ideality and the partial immiscibility of the binary HI.H 2 O and the ternary HI.I 2 .H 2 O (HIX solution) mixture make it difficult to model and simulate the dynamic behavior of the system. Therefore, it is necessary to compose separately a dynamic simulation program for Section 3 in KAERI-DySCo optimization. In this paper, a simulation program to analyze the dynamic behavior of Section 3 is introduced using the prepared KAERI-DySCo, and results of dynamic simulation are represented by running the program

Oxidation kinetics of innovative carbon materials with respect to severe air ingress accidents in HTRs and graphite disposal or processing; Oxidationskinetik innovativer Kohlenstoffmaterialien hinsichtlich schwerer Lufteinbruchstoerfaelle in HTR's und Graphitentsorgung oder Aufbereitung

Energy Technology Data Exchange (ETDEWEB)

Schloegel, Baerbel

2010-07-01

Currently future nuclear reactor concepts of the Fourth Generation (Gen IV) are under development. To some extend they apply with new, innovative materials developed just for this purpose. This thesis work aims at a concept of Generation IV Very High Temperature Reactors (VHTR) in the framework of the European project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation). The concept named ANTARES (AREVA New Technology based on advanced gas-cooled Reactors for Energy Supply) was developed by AEVA NP. It is a helium cooled, graphite moderated modular reactor for electricity and hydrogen production, by providing the necessary process heat due to its high working temperature. Particular attention is given here to oxidation kinetics of newly developed carbon materials (NBG-17) with still unknown but needed information in context of severe air ingress accident in VHTR's. Special interest is paid to the Boudouard reaction, the oxidation of carbon by CO{sub 2}. In case of an air ingress accident, carbon dioxide is produced in the primary reaction of atmospheric oxygen with reflector graphite. From there CO{sub 2} could flow into the reactor core causing further damage by conversion into CO. The purpose of this thesis is to ascertain if and to what degree this could happen. First of all oxidation kinetic data of the Boudouard reaction with NBG-17 is determined by experiments in a thermo gravimetric facility. The measurements are evaluated and converted into a common formula and a Langmuir-Hinshelwood similar oxidation kinetic equation, as input for the computer code REACT/THERMIX. This code is then applied to analyse severe air ingress accidents for several air flow rates. The results are discussed for two accident situations, in which a certain graphite burn off is achieved. All cases show much more damage to the graphite bottom reflector than to the reactor core. Thus the bottom reflector will lose its structural integrity much earlier than the

Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

Fourth Generation Reactor Concepts

International Nuclear Information System (INIS)

Furtek, A.

2008-01-01

were selected to generation IV by the GIF to further studies: Gas-Cooled Fast Reactor (GFR), Lead-Cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium-Cooled Fast Reactor (SFR), Supercritical Water-Cooled Reactor (SCWR), Very High Temperature Reactor (VHTR). These six systems would each need a dedicated effort in research and development. Some consideration for the fuel and recycling technology are common and can be shared. These common areas encompass: fuel cycles, fuels and materials choice, energy products, risk and safety, economics and proliferation and physical protection concerns.(author)

Investigation of materials for fusion power reactors

Science.gov (United States)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

The research of establishing reactor materials thermophysical properties data base

International Nuclear Information System (INIS)

Luo Danhui; Zhong Jianguo; Zhang Lili; Zhao Yongming

1992-01-01

In the process of nuclear reactor design and safety analysis, the reactor materials thermophysical properties parameters are very important as the main input data of reactor design and calculation. The goal of this work is to establish a practical, reliable data base of reactor materials thermophysical properties parameters with obvious function in reactor design, operation and safety analysis. At present phase, the focal point of this data base is to collect the materials thermophysical properties data based on the need of safety analysis in light water reactor and heavy water reactor. The materials to be chosen are as follows: Uranium, U-Al alloy, UO 2 , UO 2 -PuO 2 mixture, Zr-2, Zr-4, Zr-1% Ni alloy, Inconel-625, ZrO 2 (oxidic layer), boron carbide, cadmium in stainless steel, silver-indium-cadmium alloy, light water and heavy water, etc. The following thermophysical properties parameters are mainly included in the data base: thermal conductivity, thermal diffusivity, specific heat capacity, heat of melting, coefficient of thermal expansion, emittance, density, heat of vaporization, kinematic viscosity etc. The first phase of this work has been finished, which includes the method of establishing reactor materials thermophysical properties data base, the requirement of data collection, the requirement of establishing data base and the method of the data evaluation. This data base has been established and used on PC computer

Next Generation Nuclear Plant Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

P. E. MacDonald

2005-01-01

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen

Next Generation Nuclear Plant Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

None

2005-01-01

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: (1) Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission (2) Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented

Materials technologies of light water reactors

International Nuclear Information System (INIS)

Begley, R.

1984-01-01

Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

AGR-2 Data Qualification Interim Report

International Nuclear Information System (INIS)

Abbott, Michael L.

2010-01-01

Projects for the very high temperature reactor (VHTR) Technology Development Office program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program established the NGNP Data Management and Analysis System (NDMAS) to manage and document VHTR data qualification, for storage of the data in a readily accessible electronic form, and to assist in the analysis and presentation of the data. This document gives the status of NDMAS processing and qualification of data associated with the initial reactor cycle (147A) of the second Advanced Gas Reactor (AGR-2) experiment which began on June 21, 2010. Because it is early in the AGR-2 experiment, data from only two AGR-2 data streams are reported on: Fuel Fabrication and Fuel Irradiation data. As of August 1, 2010, approximately 311,000 irradiation data records have been stored in NDMAS, and qualification tests are in progress. Preliminary information indicates that TC 2 in Capsule 2 failed prior to start of the experiment, and NDMAS testing has thus far identified only two invalid data values from the METSO data collection system Data from the Fission Product Monitoring System (FPMS) are not currently processed until after reactor cycle shutdown and have not yet been received. A description of the ATR operating conditions data associated with the AGR-2 experiment (e.g., power levels) are summarized in the AGR-1 data qualification report (INL/EXT-09-16460). Since ATR data are collected under ATR program data quality requirements (i.e., outside the VHTR program), the NGNP program and NDMAS do not take additional actions to qualify these data other than NDMAS capture testing. Data qualification of graphite characterization data collected under the Graphite Technology Development Project is reported in a separate status report (Hull 2010).

AGR-2 Data Qualification Interim Report

Energy Technology Data Exchange (ETDEWEB)

Michael L. Abbott

2010-09-01

Projects for the very high temperature reactor (VHTR) Technology Development Office program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program established the NGNP Data Management and Analysis System (NDMAS) to manage and document VHTR data qualification, for storage of the data in a readily accessible electronic form, and to assist in the analysis and presentation of the data. This document gives the status of NDMAS processing and qualification of data associated with the initial reactor cycle (147A) of the second Advanced Gas Reactor (AGR-2) experiment which began on June 21, 2010. Because it is early in the AGR-2 experiment, data from only two AGR-2 data streams are reported on: Fuel Fabrication and Fuel Irradiation data. As of August 1, 2010, approximately 311,000 irradiation data records have been stored in NDMAS, and qualification tests are in progress. Preliminary information indicates that TC 2 in Capsule 2 failed prior to start of the experiment, and NDMAS testing has thus far identified only two invalid data values from the METSO data collection system Data from the Fission Product Monitoring System (FPMS) are not currently processed until after reactor cycle shutdown and have not yet been received. A description of the ATR operating conditions data associated with the AGR-2 experiment (e.g., power levels) are summarized in the AGR-1 data qualification report (INL/EXT-09-16460). Since ATR data are collected under ATR program data quality requirements (i.e., outside the VHTR program), the NGNP program and NDMAS do not take additional actions to qualify these data other than NDMAS capture testing. Data qualification of graphite characterization data collected under the Graphite Technology Development Project is reported in a separate status report (Hull 2010).

Stress Analysis of a TRISO Coated Particle Fuel by Using ABAQUS Finite Element Visco-Elastoplastic Solutions

International Nuclear Information System (INIS)

Cho, Moon Sung; Kim, Y. M.; Lee, Y. W.

2006-01-01

The fundamental design for a gas-cooled reactor relies on an understanding of the behavior of a coated particle fuel. KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) Project since 2004, is developing a fuel performance analysis code for a VHTR named COPA (COated Particle fuel Analysis). A validation of COPA was attempted by comparing its benchmark results with the visco-elastic solutions obtained from the ABAQUS code calculations for the IAEA-CRP-6 TRISO coated particle benchmark problems involving a creep, swelling, and pressure. However, the ABAQUS finite element model used for the above-mentioned analysis did not consider the material nonlinearity of the SiC coating layer that showed stress levels higher than the assumed yield point of the material. In this study, a consideration of the material nonlinearity is included in the ABAQUS model to obtain the visco-elastoplastic solutions and the results are compared with the visco-elastic solutions obtained from the previous ABAQUS model

Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

International Nuclear Information System (INIS)

Tak, Nam-il; Kim, Min-Hwan; Lee, Won Jae

2008-01-01

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly

The use of ferritic materials in light water reactor power plants

International Nuclear Information System (INIS)

Marston, T.V.

1984-01-01

This paper reviews the use of ferritic materials in LWR power plant components. The two principal types of LWR systems, the boiling water reactor (BWR) and the pressurized water reactor (PWR) are described. The evolution of the construction materials, including plates and forgings, is presented. The fabrication process for both reactors constructed with plates and forgings are described in detail. Typical mechanical properties of the reactor vessel materials are presented. Finally, one critical issue radiation embrittlement dealing with ferritic materials is discussed. This has been one of the major issues regarding the use of ferritic material in the construction of LWR pressure vessels

Materials development for fast reactor applications

Energy Technology Data Exchange (ETDEWEB)

Jayakumar, T.; Mathew, M.D.; Laha, K.; Sandhya, R., E-mail: san@igcar.gov.in

2013-12-15

Highlights: • A modified version of alloy D9 designated as IFAC-1 has been developed. • Oxide dispersion strengthened Grade 91 steel with good creep strength developed. • 0.14 wt% nitrogen in 316LN stainless steel leads to improved mechanical properties. • Type IV cracking resistant Grade 91 steel with boron addition developed. • Mechanical properties of SFR materials evaluated in sodium environment. -- Abstract: Materials play a crucial role in the economic competitiveness of electricity produced from fast reactors. It is necessary to increase the fuel burn-up and design life in order to realize this objective. The burnup is largely limited by the void swelling and creep resistance of the fuel cladding and wrapping materials. India's 500 MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are D9 austenitic stainless steel as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup, titanium, phosphorous and silicon contents in alloy D9 have been optimized for decreased void swelling and increased creep strength and this has led to the development of a modified version of alloy D9 as IFAC-1. Ferritic steels are inherently resistant to void swelling. The disadvantage is their poor creep strength. Creep resistance of 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long-term creep strength, comparable to alloy D9 so as to achieve higher fuel burnup. Improved versions of 316LN stainless steel with nitrogen content of about 0.14 wt% having higher creep strength to increase the life of fast reactors and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator

Detection method for nuclear reactor material

International Nuclear Information System (INIS)

Isobe, Yusuke; Hashimoto, Motoyuki.

1995-01-01

A fine state of a test piece taken out of a reactor core is analyzed upon periodical inspection, and a new test piece previously reproducing the state described above at the outside of the reactor is disposed to the reactor core upon completion of the periodical inspection. Further, a fine state of the material at a time preceding to the operation time at a certain periodical inspection is forecast, and a test piece reproducing the state at the outside of the reactor is disposed to the reactor core upon the completion of the periodical inspection. Since a test piece previously reproducing the change of the state up to a certain periodical inspection by a method other than irradiation of neutrons is newly disposed, radiation of the test piece is not extremely increased even after an extremely long period of summed up reactor operation time, to provide substantially constant radiation level on every test piece. (T.M.)

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

Application of material databases for improved reliability of reactor pressure vessels

International Nuclear Information System (INIS)

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

1994-01-01

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

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1980-01-01

The law intends under the principles of the atomic energy act to regulate the refining, processing and reprocessing businesses of nuclear raw and fuel metarials and the installation and operation of reactors for the peaceful and systematic utilization of such materials and reactors and for securing public safety by preventing disasters, as well as to control internationally regulated things for effecting the international agreements on the research, development and utilization of atomic energy. Basic terms are defined, such as atomic energy; nuclear fuel material; nuclear raw material; nuclear reactor; refining; processing; reprocessing; internationally regulated thing. Any person who is going to engage in refining businesses other than the Power Reactor and Nuclear Fuel Development Corporation shall get the special designation by the Prime Minister and the Minister of International Trade Industry. Any person who is going to engage in processing businesses shall get the particular admission of the Prime Minister. Any person who is going to establish reactors shall get the particular admission of the Prime Minister, The Minister of International Trade and Industry or the Minister of Transportation according to the kinds of specified reactors, respectively. Any person who is going to engage in reprocessing businesses other than the Power Reactor and Nuclear Fuel Development Corporation and the Japan Atomic Energy Research Institute shall get the special designation by the Prime Minister. The employment of nuclear fuel materials and internationally regulated things is defined in detail. (Okada, K.)

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

International Nuclear Information System (INIS)

Ogawa, Masuro

1989-07-01

In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

Structural materials for fusion reactors

International Nuclear Information System (INIS)

Victoria, M.; Baluc, N.; Spaetig, P.

2001-01-01

In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Selection and challenges for LFR reactor materials

Energy Technology Data Exchange (ETDEWEB)

Weisenburger, A.; Jianu, A.; Del Giacco, M.; Fetzer, R.; Heinzel, A.; Mueller, G. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Pulsed Power and Microwave Technology

2013-07-01

Nuclear energy using Fast GenIV reactors can fulfil future demands concerning CO2 free, base load capability and sustainability. One of the most promising coolants especially due to its high thermal inertia is liquid lead (Pb). Since several years researches all over the world investigate this coolant and its impact on the reactor design and by that on the materials to be selected. The LEADER project, a follow up of ELSY, aims to design a prototypical demonstrator ALFRED and to continue with several design related aspects of the ELFR reactor. For a demonstrator the criteria of material selection are somewhat different to a commercial type like the ELFR. Material selection for ELFR of course considers all the aspects relevant for ALFRED plus the targeted burn up and the expected total dpa related damage especially of the fuel pins. In the past compatibility of structural material (steels like 316L, T91 and 15-15Ti (1.4970)) that can be employed for Pb cooled fast nuclear reactors were investigated in several EU projects like EUROTRANS and worldwide. Solubility of steel alloying elements like Ni, Fe, Cr is the driving force for the reduced corrosion resistance in contact with Pb. In-situ oxidation is the acknowledged measure to protect steels in Pb up to certain temperatures that are material dependent. Based on experiments and the derived temperature limits the average core outlet temperatures of ALFRED and the ELFR are set to 480 C. The most challenging conditions with respect to temperature are at the fuel assembly and the heat exchangers. For both, thin stable oxide scales with negligible reduction in heat transfer are the requested protection method. This presentation will give an overview on the selected materials for ALFRED and ELFR considering, beside pure compatibility, the influence of mechanical interaction like creep and fretting. (orig.)

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Thermophysical properties of materials for water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA`s International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs.

Thermophysical properties of materials for water cooled reactors

International Nuclear Information System (INIS)

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

Investigation of FIV Characteristics on a Coaxial Double-tube Structure

Energy Technology Data Exchange (ETDEWEB)

Song, Kee Nam; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Sang Chul [ABLEMAX Co., Seoul (Korea, Republic of)

2009-10-15

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of 950 .deg. C for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

the JHR Material Testing Reactor

International Nuclear Information System (INIS)

Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

2013-06-01

The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

Materials surveillance program for C-E NSSS reactor vessels

International Nuclear Information System (INIS)

Koziol, J.J.

1977-01-01

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

High temperature brazing of reactor materials

International Nuclear Information System (INIS)

Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

1990-01-01

Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

Overview of environmental materials degradation in light-water reactors

International Nuclear Information System (INIS)

Shaaban, H.I.; Wu, P.

1986-08-01

This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

Characterization of 2D-C/C composite for application of very high temperature reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sumita, Junya; Kunimoto, Eiji; Sawa, Kazuhiro; Makita, Taiyo; Takagi, Takashi; Kim, W.J.; Jung, C.H.; Park, J.Y.

2010-01-01

For in-core components of VHTR (Very High Temperature Reactor), carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials. In this study, fracture behaviors of two dimensional (2D-) C/C composites were examined by SENB specimens with four-point bending test. The surface of specimens was observed by a CCD camera during the bending test, and observed by a stereomicroscope before and after the bending test. The following results were obtained through mode-I fracture test. (1) Three types of the composites were evaluated by tentatively using the stress intensity factor equation for metallic materials. The equivalent stress intensity factor of 2D-C/C composite is in the range of 5.9 - 10.0MPa m 1/2 . It was expected that the fracture mechanism for the composite materials could be assessed by this test method. (2) The crack opening displacement-load behavior of C/C composite might depend not only on the propagation of crack but also on delaminating between layers. (author)

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

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

1992-08-01

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

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

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

AGR-1 Data Qualification Interim Report

International Nuclear Information System (INIS)

Abbott, Machael

2009-01-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010

Materials for advanced reactor facilities: development and application. Materials of School-Conference for young scientists and specialists

International Nuclear Information System (INIS)

2012-01-01

In the collection of works there are the texts, summaries and presentations of lectures delivered by the leading specialists of the branch as well as the abstracts of the students of school-conference for young scientists and specialists Materials for advanced reactor facilities: development and application, which took place on October, 29 - November, 2, 2012 in Zvenigorod. In the materials presented different aspects of development and application of materials of reactor cores and vessels of advanced reactors, computerized simulation of properties of radiation-resistant materials and simulation investigations of material radiation hardness are considered [ru

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

Radiation Damage in Reactor Materials. Part of the Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-08-15

Radiation damage has presented a new design parameter for the selection of materials to be used in fuel and cladding elements, moderators, structural components and pressure vessels in nuclear reactors. The severe and novel requirements for certain optimum combinations of physical and nuclear properties have emphasized the need for a better understanding of the basic mechanisms of radiation damage. This knowledge is not only essential for progress in the field of nuclear energy, but has direct applications to space technology and semi-conductor research as well. The IAEA, as part of its programme of promoting nuclear technology, therefore convened the Symposium on Radiation Damage in Solids and Reactor Materials, 7-11 May 1962. At the invitation of, and with generous material assistance from, the Government of Italy, the Symposium was held at Venice. The Symposium was primarily concerned with the investigation of the fundamental processes of radiation that underlie the behaviour of metals, alloys and ceramics that are actually useful or potentially useful reactor materials. Two sessions were devoted to studies of irradiation effects on simple metals, as these effects are easiest to interpret. Other topics included general theory, alloys, fissionable and moderator materials and special experimental techniques for radiation damage studies. The properties influenced by irradiation which were of main concern were those of primary importance to the behaviour of solids as reactor materials (e. g. dimensional stability, phase transformation, radiation hardening, fracture, fission-gas escape from uranium and its compounds). Other properties, such as optical, electrical and magnetic properties, and effects on semiconductors, ionic and other non-metallic crystals are also of interest in that these studies can increase our knowledge of the mechanism of radiation damage in solids and provide a tool for investigation into the physics of the solid state by offering a means of

Radiation Damage in Reactor Materials. Part of the Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials

International Nuclear Information System (INIS)

1963-01-01

Radiation damage has presented a new design parameter for the selection of materials to be used in fuel and cladding elements, moderators, structural components and pressure vessels in nuclear reactors. The severe and novel requirements for certain optimum combinations of physical and nuclear properties have emphasized the need for a better understanding of the basic mechanisms of radiation damage. This knowledge is not only essential for progress in the field of nuclear energy, but has direct applications to space technology and semi-conductor research as well. The IAEA, as part of its programme of promoting nuclear technology, therefore convened the Symposium on Radiation Damage in Solids and Reactor Materials, 7-11 May 1962. At the invitation of, and with generous material assistance from, the Government of Italy, the Symposium was held at Venice. The Symposium was primarily concerned with the investigation of the fundamental processes of radiation that underlie the behaviour of metals, alloys and ceramics that are actually useful or potentially useful reactor materials. Two sessions were devoted to studies of irradiation effects on simple metals, as these effects are easiest to interpret. Other topics included general theory, alloys, fissionable and moderator materials and special experimental techniques for radiation damage studies. The properties influenced by irradiation which were of main concern were those of primary importance to the behaviour of solids as reactor materials (e. g. dimensional stability, phase transformation, radiation hardening, fracture, fission-gas escape from uranium and its compounds). Other properties, such as optical, electrical and magnetic properties, and effects on semiconductors, ionic and other non-metallic crystals are also of interest in that these studies can increase our knowledge of the mechanism of radiation damage in solids and provide a tool for investigation into the physics of the solid state by offering a means of

Interatomic potentials for fusion reactor material simulations

International Nuclear Information System (INIS)

Bjoerkas, C.

2009-01-01

In this thesis, the behaviour of a material situated in a fusion reactor was studied using molecular dynamics simulations. Simulations of processes in the next generation fusion reactor ITER include the reactor materials beryllium, carbon and tungsten as well as the plasma hydrogen isotopes. This means that interaction models, i.e. interatomic potentials, for this complicated quaternary system are needed. The task of finding such potentials is nonetheless nearly at its end, since models for the beryllium-carbon-hydrogen interactions were constructed in this thesis and as a continuation of that work, a beryllium-tungsten model is under development. These potentials are combinable with the earlier tungsten-carbon-hydrogen ones. The potentials were used to explain the chemical sputtering of beryllium due to deuterium plasma exposure. During experiments, a large fraction of the sputtered beryllium atoms were observed to be released as BeD molecules, and the simulations identified the swift chemical sputtering mechanism, previously not believed to be important in metals, as the underlying mechanism. Radiation damage in the reactor structural materials vanadium, iron and iron chromium, as well as in the wall material tungsten and the mixed alloy tungsten carbide, was also studied in this thesis. Interatomic potentials for vanadium, tungsten and iron were modified to be better suited for simulating collision cascades that are formed during particle irradiation, and the potential features affecting the resulting primary damage were identified. Including the often neglected electronic effects in the simulations was also shown to have an impact on the damage. With proper tuning of the electronphonon interaction strength, experimentally measured quantities related to ion-beam mixing in iron could be reproduced. The damage in tungsten carbide alloys showed elemental asymmetry, as the major part of the damage consisted of carbon defects. On the other hand, modelling the damage

Reference design (MK-I and MK-II) for experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Miyamoto, Yoshiaki; Suzuki, Kunihiko; Sato, Sadao

1975-10-01

This report summarizes the results of a study on thermal and mechanical performances of the core, which are obtained in course of reference design (Mk-I and Mk-II) for the experimental multi-purpose VHTR: (1) Design criteria, design methods and design data. These bases are also discussed in order to refer in the case of proceeding a next design work. (2) The results of performance analysis such as the initial core and its prediction for the irradiated core. (auth.)

Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 3

International Nuclear Information System (INIS)

G.O. Hayner; R.L. Bratton; R.E. Mizia; W.E. Windes; W.R. Corwin; T.D. Burchell; C.E. Duty; Y. Katoh; J.W. Klett; T.E. McGreevy; R.K. Nanstad; W. Ren; P.L. Rittenhouse; L.L. Snead; R.W. Swindeman; D.F. Wlson

2006-01-01

Plan (PPMP), INL/EXT-05-00952, Rev. 1, March, 2006. This document provides planning options for the development of the NGNP Project, a discussion of the project programmatic risks and a discussion of the most important deliverables to support Critical Decision-1 (CD-1) required by the DOE Acquisition Management System. This plan in conjunction with other plan and study documents to be issued will establish the detailed planning basis for the NGNP Project. Based on these guidelines and other studies performed previously, DOE has selected the Very High Temperature Reactor (VHTR) design for the NGNP Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, tri-isotopic (TRISO)-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: A full-scale prototype VHTR by about 2021; High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; Nuclear-assisted production of hydrogen with a focus on economic performance; and By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and

Gen IV Materials Handbook Implementation Plan

International Nuclear Information System (INIS)

Rittenhouse, P.; Ren, W.

2005-01-01

A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

Whole core transport calculation for the VHTR hexagonal core

International Nuclear Information System (INIS)

Cho, J. Y.; Kim, K. S.; Lee, C. C.; Joo, H. G.

2007-01-01

Recently, the DeCART code which performs the whole core calculation by coupling the radial MOC transport kernel with the axial nodal kernel has equipped a kernel to deal with the hexagonal geometry and applied to the VHTR hexagonal core to examine the accuracy and the computational efficiency of the implemented kernel. The implementation includes a modular ray tracing module based on the hexagonal assembly and a multi-group CMFD module to perform an efficient transport calculation. The requirements for the modular ray are: (1) the assembly based path linking and (2) the complete reflection capabilities. The first requirement is met by adjusting the azimuthal angle and the ray spacing for the modular ray to construct a core ray by the path linking. The second requirement is met by expanding the constructed azimuthal angle in the range of [0,30 degree] to the remained range to reflect completely at the core boundaries. The considered reflecting surface angles for the complete reflection are 30n's (n=1,2,1,12). The CMFD module performs the equivalent diffusion calculation to the radial MOC transport calculation based on the homogenized structure units. The structure units include the hexagonal pin cells and gap cells appearing at the assembly boundary. Therefore, the CMFD module is programmed to deal with the unstructured cells such as the gap cells. The CMFD equation consists of the two parts of (1) the conventional FDM and (2) the current corrective parts. Since the second part of the CMFD equation guarantees the reproducibility of the radial MOC transport solutions for the cell averaged reaction rate and the net current at the cell surfaces, how to build the first part of the CMFD equation is not important. Therefore, the first part of the CMFD equation is roughly built by using the normal distance from the gravity center to the surface. The VHTR core uses helium as a coolant which is realized as a void hole in a neutronics calculation. This void hole which

A dynamic study on the sulfuric acid distillation column for VHTR-assisted hydrogen production systems

International Nuclear Information System (INIS)

Youngjoon, Shin; Heesung, Shin; Jiwoon, Jang; Kiyoung, Lee; Jonghwa, Chang

2007-01-01

The sulfur-iodine (SI) cycle and the Westinghouse sulfur hybrid cycle coupled to a very high temperature gas-cooled reactor (VHTR) are well known as a feasible technology to produce hydrogen. The concentration of the sulfuric acid solution and its decomposition are essential parts in both cycles. In this paper, the thermophysical properties which are the boiling point, latent heat, and the partial pressures of water, sulfuric acid, and sulfur trioxide have been correlated as a function of the sulfuric acid concentration for the H 2 SO 4 and H 2 O binary chemical system, based on the data in Perry's chemical engineers' hand-book and other experimental data. By using these thermophysical correlations, a dynamic analysis of a sulfuric acid distillation column has been performed to establish the column design requirements and its optimum operation condition. From the results of the dynamic analysis, an optimized column system is anticipated for a distillation column equipped with 2 ideal plates and a second plate feeding system from the bottom plate. The effects of the hold-up of the re-boiler and the reflux ratio from the top product stream on the elapsing time when the system progresses toward a steady state have been analyzed. (authors)

Irradiation devices for fusion reactor materials results obtained from irradiated lithium aluminate at the OSIRIS reactor

International Nuclear Information System (INIS)

Lefevre, F.; Thevenot, G.; Rasneur, B.; Botter, F.

1986-06-01

Studies about controlled fusion reactor of the Tokamak type require the examination of the radiation effects on the behaviour of various potential materials. Thus, in the first part of this paper, are presented the devices adapted to these materials studies and used in the OSIRIS reactor. In a second part, is described an experiment of irradiation ceramics used as candidates for breeding material and are given the first results

Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

International Nuclear Information System (INIS)

Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

2006-01-01

Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

In-core materials testing under LWR conditions in the Halden reactor

International Nuclear Information System (INIS)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

2002-01-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

In-core materials testing under LWR conditions in the Halden reactor

Energy Technology Data Exchange (ETDEWEB)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

2002-07-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

International Nuclear Information System (INIS)

Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

2009-01-01

Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

Investigation of structural materials of reactors using high-energy heavy-ion irradiations

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

ITER at the international conference on fusion reactor materials

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.; Matera, R.

1998-01-01

The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

Tungsten-based composite materials for fusion reactor shields

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.

1985-01-01

Composite tungsten-based materials were recently proposed for the heavy constituent of compact fusion reactor shields. These composite materials will enable the incorporation of tungsten - the most efficient nonfissionable inelastic scattering (as well as good neutron absorbing and very good photon attenuating) material - in the shield in a relatively cheap way and without introducing voids (so as to enable minimizing the shield thickness). It is proposed that these goals be achieved by bonding tungsten powder, which is significantly cheaper than high-density tungsten, with a material having the following properties: good shielding ability and relatively low cost and ease of fabrication. The purpose of this work is to study the effectiveness of the composite materials as a function of their composition, and to estimate the economic benefit that might be gained by the use of these materials. Two materials are being considered for the binder: copper, second to tungsten in its shielding ability, and iron (or stainless steel), the common fusion reactor shield heavy constituent

Application of mesh free lattice Boltzmann method to the analysis of very high temperature reactor lower plenum

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Dept. of Energy and Environment

2011-11-15

Inside a helium-cooled very high temperature reactor (VHTR) lower plenum, hot gas jets from upper fuel channels with very high velocities and temperatures and is mixed before flowing out. One of the major concerns is local hot spots in the plenum due to inefficient mixing of the helium exiting from differentially heated fuel channels and it involves complex fluid flow physics. For this situation, mesh-free technique, especially Lattice Boltzmann Method (LBM), is thus of particular interest owing to its merit of no mesh generation. As an attempt to find efficiency of the method in such a problem, 3 dimensional flow field inside a scaled test model of the VHTR lower plenum is computed with commercial XFLOW code. Large eddy simulation (LES) and classical Smagorinsky eddy viscosity (EV) turbulence models are employed to investigate the capability of the LBM in capturing large scale vortex shedding. (orig.)

Development of multipurpose VHTR

International Nuclear Information System (INIS)

Ishikawa, Hiroshi; Aochi, Tetsuo; Hara, Masao

1983-01-01

In order to introduce atomic energy, which has been utilized mostly for electric power generation, into non-electric power field which amounts to 60 - 70% of energy demand in Japan, the development of a multi-purpose high temperature gas-cooled reactor has been advanced in the Japan Atomic Energy Research Institute. Including the progress and trend of the development of high temperature gas-cooled reactors in foreign countries, the features, necessity, the state of research and development and the way of thinking about heat utilization system regarding the reactors of this type are summarized. Since the Dragon Project of OECD in 1959, the course of the development of high temperature gas-cooled reactors is described. In Japan, the utilization of nuclear thermal energy for iron-making process was investigated to resolve environmental problems and to get rid of coal. It was decided to construct an experimental reactor, aiming at the start of operation around 1990. The features of high temperature gas-cooled reactors, the utilization mode of nuclear thermal energy, the design of an experimental reactor, the research and development related to the experimental reactor and the heat utilization system for the experimental reactor, the trend of development in FRG, USA and USSR are described. (Kako, I.)

AGR-1 Data Qualification Report

International Nuclear Information System (INIS)

Abbott, Michael

2010-01-01

Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: (1) Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). (2) Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. (3) FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report currently

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1979-01-01

The law aims to perform regulations on enterprises of refining, processing and reprocessing of nuclear source and fuel materials and on establishment and operation of reactors to realize the peaceful and deliberate utilization of atomic energy according to the principle of the atomic energy basic law. Regulations of use of internationally regulated substances are also envisaged to observe international agreements. Basic concepts and terms are defined, such as: atomic energy; nuclear fuel material; nuclear source material; reactor; refining; processing; reprocessing and internationally regulated substance. Any person besides the Power Reactor and Nuclear Fuel Material Developing Corporation who undertakes refining shall be designated by the Prime Minister and the Minister of International Trade and Industry. An application shall be filed to the ministers concerned, listing name and address of the person, name and location of the refining works, equipment and method of refining, etc. The permission of the Prime Minister is necessary for any person who engages in processing. An application shall be filed to the Prime Minister, listing name and address of the person, name and location of the processing works and equipment and method of processing, etc. Permission of the Prime Minister, the Minister of International Trade and Industry or the Minister of Transport is necessary for any person who sets up reactors. An application shall be filed to the minister concerned, listing name and address of the person, purpose of operation, style, thermal output of reactor and number of units, etc. (Okada, K.)

Aspects of nuclear process heat application of very high temperature reactors (VHTR)

International Nuclear Information System (INIS)

Jansing, W.T.; Kugeler, K.

2014-01-01

been tested successfully. The test time in totally was longer than 10 000 h. A hot steam generator with a power of 10 MW (T He =950°C, p He =40 bar) for application in different processes has been tested over a long time with good success too. For the steam gasification of coal or other C-containing substances a special gasifier has undergone long time testing with great success. This component (T He =950°C, p=40 bar) for application in different processes has been tested over a time of more than 10 000 h. This component (P≈ 3MW,T He =1000°C, p He =40 bar, T gasif =800°C) represented a characteristic part of a large fluidized bed gasifier. The development work additionally contained a broad material program for alloys applied at helium temperatures of around 950°C. All mechanical relevant data till 30000 hours, data of corrosion, Tritium- and hydrogen permeation have been measured for some promising candidates. Special experiments related to reaction kinetics, heat transfer, pressure drops, vibration, friction and wear, behavior on special components in helium have been carried out and delivered a broad know how on helium technologies. During the planning work for different reactor concepts all questions of coupling nuclear reactors and processes have been analyzed in detail and partly have undergone steps of a nuclear licensing process. It was shown, that extreme safety requirements of the nuclear heat source and the total plant could be fulfilled. (author)

Code qualification of structural materials for AFCI advanced recycling reactors.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

Energy Technology Data Exchange (ETDEWEB)

Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

1968-06-01

Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

CLUPH: a Fortran program of collision probabilities for hexagonal lattice and its application to VHTR

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Gotoh, Yorio

1981-02-01

A new collision probability routine CLUPH was added to the computer program set LAMP-B to analyse the hexagonal VHTR fuel and control blocks where in addition to the annular array of fuel pin rods the asymmetric insertions of burnable poison rods and control rods are characteristic. The perfect reflective boundary condition is no more realistic to consider the arrangement of asymmetric hexagonal blocks. The periodic and the rotational arrangement of blocks are surveyed to consider the interference effect between the burnable poison rods. In addition the effects of coated particle fuel in fuel rod, and of B 4 C grain in burnable poison rod, are investigated. The average cross sections of control rod block were derived from the calculation of a super cell which consists of the control rod block and of the surrounding six fuel blocks. The care was taken to the control rod block located at the core-reflector boundary by replacing a sector of surrounding material in supper cell by reflector material. The two dimensional diffusion calculations of simplified cores of Mk-III were performed to obtain the reactivity worths of control rods, for illustration. (author)

Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

International Nuclear Information System (INIS)

1978-01-01

With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

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

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.

Generation 4 International Forum. 2009 GIF R and D outlook for generation 4 nuclear energy systems

International Nuclear Information System (INIS)

2009-01-01

This document presents the state, at mid 2009, of research and development of the 6 reactor types that were selected in the framework of the GIF (Generation 4 International Forum): VHTR (Very High Temperature Reactor), SFR (Sodium-cooled Fast Reactor), SCWR (Super-Critical Water Reactor), GFR (Gas-cooled Fast Reactor), LFR (Lead-cooled reactor), and MSR (Molten Salt Reactor). Regarding each type of reactors, the state of advancement is reported for the reactor itself, its specific components and materials, its nuclear fuel, and its fuel cycle. The outlook of development and research work is also given for the next 5 years for the 6 types of reactors. (A.C.)

Material choices for the commercial fast reactor steam generators

International Nuclear Information System (INIS)

Willby, C.; Walters, J.

1978-01-01

Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

Material options for a commercial fusion reactor first wall

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m 2 . A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW 2 , provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys

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)

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

International Nuclear Information System (INIS)

Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

International Nuclear Information System (INIS)

Wood, D.S.

The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

International Nuclear Information System (INIS)

Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

1995-01-01

A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

Fusion-reactor blanket-material safety-compatibility studies

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

1982-11-01

Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

Advanced materials: The key to attractive magnetic fusion power reactors

International Nuclear Information System (INIS)

Bloom, E.E.

1992-01-01

Fusion is one of the most attractive central station power sources from the viewpoint of potential safety and environmental impact characteristics. Studies also indicate that fusion can be economically competitive with other options such as fission reactors and fossil-fired power stations. However, to achieve this triad of characteristics we must develop advanced materials with properties tailored for performance in the various fusion reactor systems. This paper discusses the desired characteristics of materials and the status of materials technology in four critical areas: (1) structural material for the first wail and blanket (FWB), (2) plasma-facing materials, (3) materials for superconducting magnets, and (4) ceramics for electrical and structural applications

Advanced materials - the key to attractive magnetic fusion power reactors

International Nuclear Information System (INIS)

Bloom, E.E.

1992-01-01

Fusion is one of the most attractive central station power sources from the viewpoint of potential safety and environmental impact characteristics. Studies also indicate that fusion can be economically competitive with other options such as fission reactors and fossil-fired power stations. However, to achieve this triad of characteristics we must develop advanced materials with properties tailored for performance in the various fusion reactor systems. This paper discusses the desired characteristics of materials and the status of materials technology in four critical areas: (1) structural materials for the first wall and blanket (FWB), (2) plasmafacing materials, (3) materials for superconducting magnets, and (4) ceramics for electrical and structural applications. (author)

Overview of fast reactor structural materials programme in India

International Nuclear Information System (INIS)

Rodriguez, P.; Paranjpe, S.R.; Chetal, S.C.; Mannan, S.L.; Ray, S.K.; Seetharaman, V.; Srinivasan, G.

The fast reactor structural materials activities in India comprise of the programme on the materials for the Fast Breeder Test Reactor (FBTR), the construction of which is nearing completion, and the programme on the candidate materials for the Prototype Fast Breeder Reactor (PFBR) which is now in the design stage. For the materials in use in FBTR, the main thrust has been towards detailed evaluation and documentation of long term (creep) properties of type 316 stainless steel base material in air. For the PFBR the philosophy has been to identify the candidate materials and to evolve a wider scope for the testing and evaluation programmes. The major structural component is identified as variants of type 304 stainless steel and the programmes undertaken include study of low cycle fatigue properties and environmental effects on creep and stress rupture properties. Evaluations of aging embrittlement of type 316 stainless steel base material and weldments are also in progress. The paper lists the testing programmes identified for adoption in the near future. These include creep-fatigue damage studies and fracture mechanics studies on weldments for type 304 stainless steel and testing programme on 2.25 Cr-1 Mo and 9 Cr-1 Mo steels, the identified candidate materials for steam generators. The development efforts also include a comprehensive programme on inelastic analysis procedure. (author)

Materials science research for sodium cooled fast reactors

Indian Academy of Sciences (India)

The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

Numerical solution of heat transfer process in a prismatic VHTR core accompanying bypass and cross flows

International Nuclear Information System (INIS)

Wang, Li; Liu, Qiusheng; Fukuda, Katsuya

2016-01-01

Highlights: • Three-dimensional CFD analysis is conducted for the thermal analysis in the reactor core. • Hot spot temperature, coolant channel outlet temperature distribution are affected by bypass flow. • Bypass gap size has significant influence on temperature and flow distribution in the core. • Cross flow has some effect on the temperature distribution of the coolant in the core due to flow mixing in the cross gaps. - Abstract: Bypass flow and cross flow gaps both exist in the core of a very high temperature gas-cooled reactor (VHTR), which is inevitable owing to tolerances in manufacturing, thermal expansion and irradiation shrinkage. The coolant mass flow rate distribution, temperature distribution, and hot spot temperature are significantly affected by bypass and cross flows. In the present study, three-dimensional CFD analysis is conducted for thermal analysis of the reactor core. A validation study for the turbulence model is performed by comparing the friction coefficient with published correlations. A sensitivity study of the near wall mesh is conducted to ensure mesh quality. Parametric studies are performed by changing the size of the bypass and cross gaps using a one-twelfth sector of a fuel block. Simulation results show the influence of the bypass gap size on temperature distribution and coolant mass flow rate distribution in the prismatic core. It is shown that the maximum fuel and coolant channel outlet temperatures increase with an increase in the gap size, which may lead to a structural risk to the fuel block. The cross flow is divided into two types: the cross flow from the bypass gap to the coolant channels and the cross flow from the high-pressure coolant channels to low-pressure coolant channels. These two types of flow have an opposing influence on the temperature gradient. It is found that the presence of the cross flow gaps may have a significant effect on the distribution of the coolant in the core due to flow mixing in the

Design and fabrication of irradiation testing capsule for research reactor materials

International Nuclear Information System (INIS)

Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

2012-01-01

Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

Design and fabrication of irradiation testing capsule for research reactor materials

Energy Technology Data Exchange (ETDEWEB)

Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

Metallic materials corrosion problems in molten salt reactors

International Nuclear Information System (INIS)

Chauvin, G.; Dixmier, J.; Jarny, P.

1977-01-01

The USA forecastings concerning the molten salt reactors are reviewed (mixtures of fluorides containing the fuel, operating between 560 and 700 0 C). Corrosion problems are important in these reactors. The effects of certain characteristic factors on corrosion are analyzed: humidity and metallic impurities in the salts, temperature gradients, speed of circulation of salts, tellurium from fission products, coupling. In the molten fluorides and experimental conditions, the materials with high Ni content are particularly corrosion resistant alloys (hastelloy N). The corrosion of this material is about 2.6 mg.cm -2 at 700 0 C [fr

Non-proliferation issues for the disposition of fissile materials using reactor alternatives

International Nuclear Information System (INIS)

Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

1996-01-01

The Department of Energy (DOE) is analyzing long-term storage on options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to bum the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the nonproliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

Non-proliferation issues for the disposition of fissile materials using reactor alternatives

International Nuclear Information System (INIS)

Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

1996-01-01

The Department of Energy (DOE) is analyzing long-term storage imposition options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to burn the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the non-proliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

Raw materials for reflector graphite (for reactors)

International Nuclear Information System (INIS)

Wilhelmi, G.; Mindermann, D.

1992-01-01

The manufacturing concept for the core components of German high temperature reactor (HTR) types of graphite was previously entirely directed to the use of German tar coke (St coke). As the plants for producing this material no longer complied technically with the current environmental protection requirements, one had to assume that they would soon be shut down. To prevent bottlenecks in the erection of future HTR plants, alternative cokes produced by modern processes by Japanese manufacturers were checked for their suitability for the manufacture of reactor graphite. This report describes the investigations carried out on these materials from the safe delayed coking process. The project work, apart from analysis of the main data of the candidate coke considered, included the processing of the raw materials into directly and secondarily extruded graphite rods on the laboratory scale, including characterisation. As the results show, the material data achieved with the previous raw material can be reproduced with Japanese St coke. The tar coke LPC-A from the Nippon Steel Chemical Co., Ltd was decided on as the new standard coke for manufacturing reflector graphite. (orig.) With 15 tabs., 2 figs [de

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

Energy Technology Data Exchange (ETDEWEB)

Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O₂ cannot be ignored, especially for the FHR, in which environment the product, SiO₂, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

International Nuclear Information System (INIS)

Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

1977-02-01

Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

Corrosion of Selected Materials in Boiling Sulfuric Acid for the Nuclear Power Industries

International Nuclear Information System (INIS)

Kim, Dong Jin; Lee, Han Hee; Kwon, Hyuk Chul; Kim, Hong Pyo; Hwang, Seong Sik

2007-01-01

Iodine sulfur (IS) process is one of the promising processes for a hydrogen production by using a high temperature heat generated by a very high temperature gas cooled reactor(VHTR) in the nuclear power industries. Even though the IS process is very efficient for a hydrogen production and it is not accompanied by a carbon dioxide evolution, the highly corrosive environment of the process limits its application in the industry. Corrosion tests of selected materials were performed in sulfuric acid to select appropriate materials compatible with the IS process. The materials used in this work were Fe-Cr alloys, Fe-Ni-Cr alloys, Fe-Si alloys, Ni base alloys, Ta, Ti, Zr, SiC, Fe-Si, etc. The test environments were 50 wt% sulfuric acid at 120 .deg. C and 98 wt% at 320 .deg. C. Corrosion rates were measured by using a weight change after an immersion. The surface morphologies and cross sectional areas of the corroded materials were examined by using SEM equipped with EDS. Corrosion behaviors of the materials were discussed in terms of the chemical composition of the materials, a weight loss, the corrosion morphology, the precipitates in the microstructure and the surface layer composition

Damage analysis and fundamental studies for fusion reactor materials development

International Nuclear Information System (INIS)

Odette, G.R.; Lucas, G.E.

1991-09-01

The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base (expected to be largely fission reactor based) to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Present status of Japan materials testing reactor

Energy Technology Data Exchange (ETDEWEB)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Present status of Japan materials testing reactor

International Nuclear Information System (INIS)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

2012-01-01

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

Science.gov (United States)

Murty, K. L.; Charit, I.

2008-12-01

Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

Graphites and composites irradiations for gas cooled reactor core structures

International Nuclear Information System (INIS)

Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

2008-01-01

Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

Materials research with neutron beams from a research reactor

Energy Technology Data Exchange (ETDEWEB)

Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

2015-03-15

Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

Compression device for feeding a waste material to a reactor

Science.gov (United States)

Williams, Paul M.; Faller, Kenneth M.; Bauer, Edward J.

2001-08-21

A compression device for feeding a waste material to a reactor includes a waste material feed assembly having a hopper, a supply tube and a compression tube. Each of the supply and compression tubes includes feed-inlet and feed-outlet ends. A feed-discharge valve assembly is located between the feed-outlet end of the compression tube and the reactor. A feed auger-screw extends axially in the supply tube between the feed-inlet and feed-outlet ends thereof. A compression auger-screw extends axially in the compression tube between the feed-inlet and feed-outlet ends thereof. The compression tube is sloped downwardly towards the reactor to drain fluid from the waste material to the reactor and is oriented at generally right angle to the supply tube such that the feed-outlet end of the supply tube is adjacent to the feed-inlet end of the compression tube. A programmable logic controller is provided for controlling the rotational speed of the feed and compression auger-screws for selectively varying the compression of the waste material and for overcoming jamming conditions within either the supply tube or the compression tube.

Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

International Nuclear Information System (INIS)

Boer, Brian

2007-01-01

The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Reliability Evaluation on Creep Life Prediction of Alloy 617 for a Very High Temperature Reactor

International Nuclear Information System (INIS)

Kim, Woo-Gon; Hong, Sung-Deok; Kim, Yong-Wan; Park, Jae-Young; Kim, Seon-Jin

2012-01-01

This paper evaluates the reliability of creep rupture life under service conditions of Alloy 617, which is considered as one of the candidate materials for use in a very high temperature reactor (VHTR) system. A Z-parameter, which represents the deviation of creep rupture data from the master curve, was used for the reliability analysis of the creep rupture data of Alloy 617. A Service-condition Creep Rupture Interference (SCRI) model, which can consider both the scattering of the creep rupture data and the fluctuations of temperature and stress under any service conditions, was also used for evaluating the reliability of creep rupture life. The statistical analysis showed that the scattering of creep rupture data based on Z-parameter was supported by normal distribution. The values of reliability decreased rapidly with increasing amplitudes of temperature and stress fluctuations. The results established that the reliability decreased with an increasing service time.

Fusion reactor materials semiannual progress report for period ending September 30, 1990

International Nuclear Information System (INIS)

1991-04-01

This is the ninth in series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development of Irradiation Performance; Damage Analysis and Fundamental Studies; and Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

Fusion reactor materials: Semiannual progress report for period ending September 30, 1987

International Nuclear Information System (INIS)

1988-03-01

This is the third in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development for Irradiation Performances; Damage Analysis and Fundamental Studies; Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

International Expert - OECD/NEA

International Nuclear Information System (INIS)

Yi, Yong Sun

2009-11-01

This report was prepared to describe the activities of Yongsun Yi as a technical secretary for the Generation IV VHTR (Very High Temperature Reactor) system. The contents of the report are; i) the GIF (Generation IV International Forum); ii) the GIF governance and technical secretariat; iii) the brief description of the VHTR system iv) the activities of Yongsun Yi as the technical secretary for the VHTR system

Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1985-01-01

This ordinance is stipulated under the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors. The designation for refining and processing businesses under the law shall be obtained for each works or enterprise where these operations are to be practiced. Persons who intend to accept the designation shall file applications attaching business plans and the other documents specified by the ordinances of the Prime Minister's Office and other ministry orders. The permission for the installation of nuclear reactors under the law shall be received for each works or enterprise where reactors are to be set up. Persons who intend to get the permission shall file applications attaching the financing plans required for the installation of reactors and the other documents designated by the orders of the competent ministry. The permission concerning the reactors installed on foreign ships shall be obtained for each ship which is going to enter into the Japanese waters. Persons who ask for the permission shall file applications attaching the documents which explain the safety of reactor facilities and the other documents defined by the orders of the Ministry of Transportation. The designation for reprocessing business and the application for it are provided for, respectively. The usage of nuclear fuel materials, nuclear raw materials and internationally regulated goods is ruled in detail. (Kubozone, M.)

Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1981-01-01

This ordinance is stipulated under the law concerning the regulation of nuclear raw materials, nuclear fuel materials and reactors. The designation for refining and processing businesses under the law shall be obtained for each works or enterprise where these operations are to be practiced. Persons who intend to accept the designation shall file applications attaching business plans and the other documents specified by the ordinances of the Prime Minister's Office and other ministry orders. The permission for the installation of nuclear reactors under the law shall be received for each works or enterprise where reactors are to be set up. Persons who intend to get the permission shall file applications attaching the financing plans required for the installation of reactors and the other documents designated by the orders of the competent ministry. The permission concerning the reactors installed on foreign ships shall be obtained for each ship which is going to enter into the Japanese waters. Persons who ask for the permission shall file applications attaching the documents which explain the safety of reactor facilities and the other documents defined by the orders of the Ministry of Transportation. The designation for reprocessing business and the application for it are provided for, respectively. The usage of nuclear fuel materials, nuclear raw materials and internationally regulated goods is ruled in detail.(Okada, K.)

Long-lived activation products in reactor materials

International Nuclear Information System (INIS)

Evans, J.C.; Lepel, E.L.; Sanders, R.W.; Wilkerson, C.L.; Silker, W.; Thomas, C.W.; Abel, K.H.; Robertson, D.R.

1984-08-01

The purpose of this program was to assess the problems posed to reactor decommissioning by long-lived activation products in reactor construction materials. Samples of stainless steel, vessel steel, concrete, and concrete ingredients were analyzed for up to 52 elements in order to develop a data base of activatable major, minor, and trace elements. Large compositional variations were noted for some elements. Cobalt and niobium concentrations in stainless steel, for example, were found to vary by more than an order of magnitude. A thorough evaluation was made of all possible nuclear reactions that could lead to long lived activation products. It was concluded that all major activation products have been satisfactorily accounted for in decommissioning planning studies completed to date. A detailed series of calculations was carried out using average values of the measured compositions of the appropriate materials to predict the levels of activation products expected in reactor internals, vessel walls, and bioshield materials for PWR and BWR geometries. A comparison is made between calculated activation levels and regulatory guidelines for shallow land disposal according to 10 CFR 61. This analysis shows that PWR and BWR shroud material exceeds the Class C limits and is, therefore, generally unsuitable for near-surface disposal. The PWR core barrel material approaches the Class C limits. Most of the remaining massive components qualify as either Class A or B waste with the bioshield clearly Class A, even at the highest point of activation. Selected samples of activated steel and concrete were subjected to a limited radiochemical analysis program as a verification of the computer model. Reasonably good agreement with the calculations was obtained where comparison was possible. In particular, the presence of 94 Nb in activated stainless steel at or somewhat above expected levels was confirmed

Damage by radiation in structural materials of BWR reactor vessels