Small Reactors with Simplified Design
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A Low Cost Supercritical Nuclear + Coal 3.0 Gwe Power Plant
n. 461 August 2012 A low cost supercritical Nuclear + Coal 3.0 Gwe power plant Pedro Cosme Costa Vieira 1 1 FEP-UP, School of Economics and Management, University of Porto A low cost supercritical Nuclear + Coal 3.0 Gwe power plant. Pedro Cosme Costa Vieira, [email protected] Faculdade de Economia do Porto, R. Dr. Roberto Frias, 80, 4200-464 Porto, Portugal Abstract: The rapid growth in the consumption of electricity in China and India has been covered at 80% by coal, which has the side effect of emitting CO2 to the atmosphere. The alternative is the use of nuclear energy that, to become unquestionably competitive, must use supercritical water as coolant. The problem is that the inflow temperature of a supercritical turbine must be at least 500 ºC that is impossible to attain using the mainstream PWR and it is uneconomical to accomplish in a pressure tubes similar to the CANDU and the RBMK reactors. In this paper, I propose a simple but important innovation that is the coupling of the nuclear reactor to a coal fired heater. With this apparently small improvement, it becomes feasible to build a multi-tube slightly supercritical pressure light water reactor where the outflow low temperature of the nuclear reactor, ≤ 400 ºC, results in a simple nuclear reactor. The design I propose will originate a 50% decrease in the cost of the electricity produced using nuclear energy. Keywords: Electricity production, Supercritical water reactor, Nuclear energy, Coal fired plant, Multi-tube reactor, SCWR. 1. Introduction. Every candidate to the Miss Universe Contest whishes to contribute for the end of war and starvation all over the World, above all, when child are the victims. -
Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment Volume 2, Detailed Analysis
Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within A Factory Environment Volume 2, Detailed Analysis Xuan Chen, Arnold Kotlyarevsky, Andrew Kumiega, Jeff Terry, and Benxin Wu Illinois Institute of Technology Stephen Goldberg and Edward A. Hoffman Argonne National Laboratory August 2013 This study continued the work, supported by the Department of Energy’s Office of Nuclear Energy, regarding the economic analysis of small modular reactors (SMRs). The study team analyzed, in detail, the costs for the production of factory-built components for an SMR economy for a pressurized-water reactor (PWR) design. The modeling focused on the components that are contained in the Integrated Reactor Vessel (IRV). Due to the maturity of the nuclear industry and significant transfer of knowledge from the gigawatt (GW)-scale reactor production to the small modular reactor economy, the first complete SMR facsimile design would have incorporated a significant amount of learning (averaging about 80% as compared with a prototype unit). In addition, the order book for the SMR factory and the lot size (i.e., the total number of orders divided by the number of complete production runs) remain a key aspect of judging the economic viability of SMRs. Assuming a minimum lot size of 5 or about 500 MWe, the average production cost of the first-of-the kind IRV units are projected to average about 60% of the a first prototype IRV unit (the Lead unit) that would not have incorporated any learning. This cost efficiency could be a key factor in the competitiveness of SMRs for both U.S. -
Full Core Conversion and Operational Experience with LEU Fuel of the DALAT Nuclear Research Reactor
RERTR 2014 – 35TH INTERNATIONAL MEETING ON REDUCED ENRICHMENT FOR RESEARCH AND TEST REACTORS OCTOBER 12-16, 2014 IAEA VIENNA INTERNATIONAL CENTER VIENNA, AUSTRIA Full Core Conversion and Operational Experience with LEU Fuel of the DALAT Nuclear Research Reactor NGUYEN, Nhi Dien – LUONG, Ba Vien – LE, Vinh Vinh – HUYNH, Ton Nghiem – NGUYEN, Kien Cuong Reactor Center - Dalat Nuclear Research Institute – Vietnam Atomic Energy Institute 01 Nguyen Tu Luc St, P.O.Box 61100, Dalat, Lamdong – Vietnam ABSTRACT After successful in full core conversion from HEU to LEU fuel at the end of 2011, safe operation and utilization of the Dalat Nuclear Research Reactor (DNRR) was emphasized and improved. During carrying out commissioning program for reactor start up, reactor characteristics parameters were measured and compared with design calculated data. Good agreement between experimental data and calculated data was archived and the working core has met all safety and exploiting requirements. Fuel and in-core management of LEU core were implemented by using MCNP-REBUS linkage computers code system with visual interface to make friendly using. The neutron trap was modified to serve for producing I-131 by increasing 6 containers. Other irradiation channels inside the reactor core and horizontal beam tubes are being effectively used for neutron activation analysis, fundamental research, nuclear data measurement, neutron radiography and nuclear structure study. 1. Introduction Physics and energy start-up of the Dalat Nuclear Research Reactor (DNRR) for full core conversion to low enriched uranium (LEU) fuel were performed from November 24th, 2011 until January 13th, 2012. The program provides specific instructions for manipulating fuel assemblies (FAs) loading in the reactor core and denotes about procedures for carrying out measurements and experiments during physics and energy start-up stages to guarantee that loaded LEU FAs in the reactor core are in accordance with calculated loading diagram and implementation necessary measurements to ensure for safety operation of DNRR. -
Advances in Small Modular Reactor Technology Developments
Advances in Small Modular Reactor Technology Developments Advances in Small Modular Reactor Technology Developments Technology in Small Modular Reactor Advances A Supplement to: IAEA Advanced Reactors Information System (ARIS) 2018 Edition For further information: Nuclear Power Technology Development Section (NPTDS) Division of Nuclear Power IAEA Department of Nuclear Energy International Atomic Energy Agency Vienna International Centre PO Box 100 1400 Vienna, Austria Telephone: +43 1 2600-0 Fax: +43 1 2600-7 Email: [email protected] Internet: http://www.iaea.org Printed by IAEA in Austria September 2018 18-02989E ADVANCES IN SMALL MODULAR REACTOR TECHNOLOGY DEVELOPMENTS 2018 Edition A Supplement to: IAEA Advanced Reactors Information System (ARIS) http://aris.iaea.org DISCLAIMER This is not an official IAEA publication. The material has not undergone an official review by the IAEA. The views expressed do not necessarily reflect those of the International Atomic Energy Agency or its Member States and remain the responsibility of the contributors. Although great care has been taken to maintain the accuracy of information contained in this publication, neither the IAEA nor its Member States assume any responsibility for consequences which may arise from its use. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. -
Scaling Analysis of the Osu High Temperature Test Facility During a Pressurized Conduction Cooldown Event Using Relap5- 3D
AN ABSTRACT OF THE THESIS OF Juan A. Castañeda for the degree of Master of Science in Nuclear Engineering presented on May 21, 2014. Tittle: SCALING ANALYSIS OF THE OSU HIGH TEMPERATURE TEST FACILITY DURING A PRESSURIZED CONDUCTION COOLDOWN EVENT USING RELAP5- 3D Abstract approved: Brian G. Woods In early 2000, the Generation IV International Forum (GIF) was created to perform research and development for the next generation nuclear systems. Among the selected nuclear systems was the Very High Temperature Gas-Cooled Reactor (VHTR). Then in 2008, the U.S. Department of Energy (DOE) decided that the Next Generation Nuclear Plant (NGNP) would be the VHTR. The VHTR was chosen because it has the capability to produce electricity, hydrogen and may be used for other high-temperature process heat applications. In support of licensing and validation of the VHTR, Oregon State University was tasked to develop a high temperature apparatus that will be able to capture the thermal fluids phenomena of the VHTR and perform integral effects tests for validation of existing safety codes. The design has been called the High Temperature Test Facility (HTTF), which is a scaled design of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The objective of this study was to investigate the ability of the HTTF to simulate the Pressurized Conduction Cooldown (PCC) Event during the natural circulation phase in the MHTGR. This was achieved with the aid of a thermal hydraulic systems code, RELAP5-3D. ©Copyright by Juan A. Castañeda May 21, 2014 All Rights Reserved SCALING ANALYSIS OF THE OSU HIGH TEMPERATURE TEST FACILITY DURING A PRESSURIZED CONDUCTION COOLDOWN EVENT USING RELAP5-3D by Juan A. -
Section 3.9.5, "Reactor Pressure Vessel Internals," Revision 4
NUREG-0800 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN 3.9.5 REACTOR PRESSURE VESSEL INTERNALS REVIEW RESPONSIBILITIES Primary - Organization responsible for mechanical engineering reviews Secondary - None I. AREAS OF REVIEW Reactor pressure vessel (RPV) internals consist of all structural and mechanical elements inside the reactor vessel in nuclear power plants. The U.S. Nuclear Regulatory Commission (NRC) regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities,” Appendix A, “General Design Criteria (GDC) for Nuclear Plants,” GDC 1, “Quality Standards and Records,” GDC 2, “Design Bases for Protection against Natural Phenomena,” GDC 4, "Environmental and Dynamic Effects Design Bases," and GDC 10, "Reactor Design,"; 10 CFR 50.55a, “General Provisions”; and 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants”; require that structures and components important to safety be constructed and tested to quality standards commensurate with the importance of the safety functions performed and designed with appropriate margins to withstand effects of anticipated operational occurrences and normal operation, natural phenomena such as earthquakes, postulated accidents including loss-of-coolant accidents (LOCAs), and events and conditions outside the nuclear power unit. Revision 4 – March 2017 USNRC STANDARD REVIEW PLAN This Standard Review Plan (SRP), NUREG-0800, has been prepared to establish criteria that the U.S. Nuclear Regulatory Commission (NRC) staff responsible for the review of applications to construct and operate nuclear power plants intends to use in evaluating whether an applicant/licensee meets the NRC's regulations. The SRP is not a substitute for the NRC's regulations, and compliance with it is not required. -
Integral Pressurized Water Reactor Simulator Manual
65 @ Integral Pressurized Water Reactor Simulator Manual Simulator Reactor Pressurized Water Integral Integral Pressurized Water Reactor Simulator Manual TRAINING COURSE SERIES TRAINING ISSN 1018-5518 VIENNA, 2017 TRAINING COURSE SERIES 65 TCS-65_cov_+Nr.indd 1,3 2017-05-24 12:15:04 INTEGRAL PRESSURIZED WATER REACTOR SIMULATOR MANUAL The following States are Members of the International Atomic Energy Agency: AFGHANISTAN GEORGIA OMAN ALBANIA GERMANY PAKISTAN ALGERIA GHANA PALAU ANGOLA GREECE PANAMA ANTIGUA AND BARBUDA GUATEMALA PAPUA NEW GUINEA ARGENTINA GUYANA PARAGUAY ARMENIA HAITI PERU AUSTRALIA HOLY SEE PHILIPPINES AUSTRIA HONDURAS POLAND AZERBAIJAN HUNGARY PORTUGAL BAHAMAS ICELAND QATAR BAHRAIN INDIA REPUBLIC OF MOLDOVA BANGLADESH INDONESIA ROMANIA BARBADOS IRAN, ISLAMIC REPUBLIC OF RUSSIAN FEDERATION BELARUS IRAQ RWANDA BELGIUM IRELAND SAN MARINO BELIZE ISRAEL SAUDI ARABIA BENIN ITALY SENEGAL BOLIVIA, PLURINATIONAL JAMAICA SERBIA STATE OF JAPAN SEYCHELLES BOSNIA AND HERZEGOVINA JORDAN SIERRA LEONE BOTSWANA KAZAKHSTAN SINGAPORE BRAZIL KENYA SLOVAKIA BRUNEI DARUSSALAM KOREA, REPUBLIC OF SLOVENIA BULGARIA KUWAIT SOUTH AFRICA BURKINA FASO KYRGYZSTAN SPAIN BURUNDI LAO PEOPLE’S DEMOCRATIC SRI LANKA CAMBODIA REPUBLIC SUDAN CAMEROON LATVIA SWAZILAND CANADA LEBANON SWEDEN CENTRAL AFRICAN LESOTHO SWITZERLAND REPUBLIC LIBERIA SYRIAN ARAB REPUBLIC CHAD LIBYA TAJIKISTAN CHILE LIECHTENSTEIN THAILAND CHINA LITHUANIA THE FORMER YUGOSLAV COLOMBIA LUXEMBOURG REPUBLIC OF MACEDONIA CONGO MADAGASCAR TOGO COSTA RICA MALAWI TRINIDAD AND TOBAGO CÔTE D’IVOIRE -
Reactor Coolant System
NUCLEAR SYSTEMS ENGINEERING Cho, Hyoung Kyu Department of Nuclear Engineering Seoul National University 4월6일보 강가? 능 Contents of Lecture Contents of lecture Chapter 1Principal Characteristics of Power Reactors Nuclear system . Will be replaced by the lecture note . Introduction to Nuclear Systems . CANDU . BWR and Fukushima accident . PWR (OPR1000 and APR1400) Chapter 4 Transport Equations for Single‐Phase Flow (up to energy equation) Chapter 6Thermodynamics of Nuclear Energy Conversion Systems: Nonflow and Steady Flow : First‐ and Second‐Law Applications Chapter 7 Thermodynamics of Nuclear Energy Conversion Systems : Nonsteady Flow First Law Analysis Thermodynamics Chapter 3 Reactor Energy Distribution Heat transport Chapter 8 Thermal Analysis of Fuel Elements Conduction heat transfer 3. PRESSURIZED WATER REACTOR References Lecture note 한수원 계통교재 Contents History of PWR Plant Overall Reactor Coolant System Steam and Power Conversion System Auxiliary System Plant Protection System Other systems History of PWR 1939: Nuclear fission was discovered. 1942: The world’s first chain reaction Achieved by the Manhattan Project (1942.12.02) 1951: Electricity was first generated from nuclear power EBR (Experimental Breeder Reactor‐1), Idaho, USA Power: 1.2 MWt, 200 kWe History of PWR 1950s: R&D USA: light water reactor . PWR: Westinghouse . BWR: General Electric (GE) USSR: graphite moderated light water reactor and WWER (similar to PWRs) UK and France: natural uranium fueled, graphite moderated, gas cooled reactor Canada: natural uranium fueled, heavy water reactor. 1954: initial success USSR: 5 MWe graphite moderated, light water cooled reactor was connected to grid. USA: the first nuclear submarine, the Nautilus (PWR) History of PWR 1956~1959 UK: Calder Hall‐1, 50 MWe GCR (1956) USA: Shippingport, 60 MWe PWR (1957) France: G‐2, 38 MWe GCR (1959) USSR: the ice breaker Lenin (1959) . -
Neutron Kinetics Model Development for VVER-1000 Reactor
Development of coupled thermohydraulics-neutron kinetics models with RELAP5-3D code for VVER reactors A. Shkarupa, N. Trofimova, I. Kadenko, A. Kharitonov, R. Yermolenko, T. Galatyuk Taras Shevchenko National University of Kyiv Sixth International Information Exchange Forum “Safety Analysis for NPPs of VVER and RBMK Types” 08-12 of April, 2002, Kyiv, Ukraine Introduction RELAP5-3D code models for Rivne NPP unit 1 (VVER-440), Zaporizhzhya NPP unit 5 (VVER- 1000) and Kozloduy NPP unit 6 (VVER-1000) have been developed within the framework of the Ukraine VVER Special Transient Analysis project [1]. For the development of RELAP5-3D VVER reactor models the corresponding RELAP5/Mod3.2 code models, developed and validated under the US DOE International Nuclear Safety Program, were used. One-dimensional models of the reactor pressure vessel and its internals have been replaced with their three-dimensional models for original complete plant models. More detailed description of RELAP5-3D VVER models is presented in this paper. RELAP5-3D code [2] has multi-dimensional thermal-hydraulic and neutron kinetics modeling capabilities. This removes any restrictions with respect to applicability of this code to full scope operational transients. Besides RELAP5-3D users may activate a new set of CHF correlations based on data available in the Czech Republic data bank [3]. These correlations replace the “CHF Table Look-up” method. Differences in the output of the PG and the table lookup method were estimated. One of the main purposes of the project is the application of the multi-dimensional coupled thermohydraulics-neutron kinetics models to analyzing the specific VVER transients in support of the In- Depth Safety Analysis (ISA) projects in Ukraine. -
Scaling Study of the Depressurized Conduction Cooldown Event in the High Temperature Test Facility Using RELAP5-3D/ATHENA
AN ABSTRACT OF THE THESIS OF Robert J. Aldridge for the degree of Master of Science in Nuclear Engineering presented on May 29, 2013. Title: Scaling Study of the Depressurized Conduction Cooldown Event in the High Temperature Test Facility Using RELAP5-3D/ATHENA Abstract approved: ________________________________________________________________________ Brian G. Woods In 2008, the Department of Energy (DOE) and Nuclear Regulatory Commission (NCR) decided that the Next Generation Nuclear Plant (NGNP) would be the Very High Temperature Reactor (VHTR). In support of the licensing and validation effort of the VHTR, Oregon State University was tasked with designing, building, and operating a thermal hydraulic test facility, the High Temperature Test Facility (HTTF). It is expected that the HTTF will provide data on the systemic response of the Modular High Temperature Gas Reactor (MHTGR) during the Depressurized Conduction Cooldown (DCC) event to validate safety analysis codes and the safety concept of the VHTR. The objective of this thesis was to determine the ability of the HTTF to model the DCC event compared to the MHTGR. To do this, a thermal hydraulic systems code, RELAP5-3D/ATHENA, was used to compare the systemic response of the MHTGR and HTTF during the DCC event. Two phases of the DCC event were modeled, the molecular diffusion phase and natural circulation phase. Each phase was modeled as a Separate Effects Test (SET) and as an Integral Effects Test (IET). For each test, the radial temperature profile, peak fuel temperature as a function of time, time to peak fuel temperature, air front velocity, time to onset of natural circulation, and natural circulation flow rates were compared to the MHTGR response. -
Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy
INL/EXT-08-13951 Revision 0 Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy R. E. Mizia April 2008 DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. INL/EXT-08-13951 Revision 0 Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy R. E. Mizia April 2008 Idaho National Laboratory Idaho Falls, Idaho 83415 Prepared for the U.S. Department of Energy Assistant Secretary for the Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517 EXECUTIVE SUMMARY The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 850 to 950°C and a plant design service life of 60 years. -
Preliminary Results of the Seventh Three-Dimensional Aer Dynamic Benchmark Problem Calculation
PRELIMINARY RESULTS OF THE SEVENTH THREE-DIMENSIONAL AER DYNAMIC BENCHMARK PROBLEM CALCULATION. SOLUTION WITH DYN3D AND RELAP5-3D CODES. Marek Beník, Jan Hádek Nuclear Research Institute ež plc, 250 68 ež, Czech Republic E-mail: [email protected] ABSTRACT The paper gives a brief survey of the 7th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI ež. This benchmark was defined at the 20th AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a VVER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The VVER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D . The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP th Dukovany was adopted for the 7 AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model.