The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

scholarly journals Prospects of diversification and ensuring energy safety of Uzbekistan

2019 ◽  
Vol 139 ◽  
pp. 01002 ◽  
Author(s):  
Kahraman Allaev ◽  
Tokhir Makhmudov
Keyword(s):  
Renewable Energy ◽  
Energy Balance ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Energy Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Current State ◽  
The Republic

The data on the current state of energy in Uzbekistan are given. The need to diversify the structure of the energy balance of the republic is shown, which ensures the energy security of the state in the medium and long term. It is argued that the construction of a nuclear power plant in Uzbekistan is not only expedient, but also necessary. In the future, renewable energy and nuclear power plants will become the basis of energy in Uzbekistan.

The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors

2014 ◽  
Author(s):  
Jozef Molnar ◽  
Radim Vocka
Keyword(s):  
Surveillance System ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Technical Specification ◽  
Training Simulator ◽  
History Of ◽  
Monitoring And Surveillance ◽  
Technical Specifications ◽  
European Nuclear

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system. After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016. This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

scholarly journals Contributions of the ORNL Piping Program to Nuclear Piping Design Codes and Standards

1977 ◽  
Vol 99 (1) ◽  
pp. 224-230
Author(s):  
S. E. Moore
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Manufacturing Industry ◽  
Stress Index ◽  
Paper Briefly ◽  
Technical Literature ◽  
Index Method ◽  
Engineering Data ◽  
History Of ◽  
Future Work

Since 1967, ORNL Piping Program has been engaged in providing information for the development of stress indices to be used in the analysis of piping components for nuclear power plants. This effort has surveyed the piping manufacturing industry and analyzed that industry’s products; has combed the technical literature for pertinent engineering data; has performed theoretical and experimental analysis of nuclear piping components; and has defined, tested, and improved indices for the stress-index method of analysis for piping components. This paper briefly reviews the history of piping-analysis standards; outlines the philosophy of the stress-index method of analysis; and explains some of the specific contributions made by the ORNL program to the Codes and Standards. Current and future work is also noted.

Setup of the Supercritical CO2 Test Facility “SCARLETT” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger
Keyword(s):  
Heat Exchanger ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

Experimental Investigation of Steam Bubble Condensation in Flowing Subcooled Water With Two Different Injection Nozzle Geometries

2013 ◽  
Author(s):  
Suleiman Al Issa ◽  
Patricia B. Weisensee
Keyword(s):  
Test Section ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Bubble Diameter ◽  
Large Diameter ◽  
Low Pressure ◽  
Test Facility ◽  
Experimental Investigations ◽  
Subcooled Water

A multiphase flow test facility was built at the Department of Nuclear Engineering at the Technical University Munich. The main goal of this facility is to investigate the condensation of steam bubbles injected into a vertical large diameter pipe (104 mm) with flowing subcooled water (6–15 K) at low pressure conditions (1.1–1.45 bar). Current experimental investigations will contribute to a better understanding of subcooled boiling at low pressures, accidental conditions in nuclear power plants and low-pressure research reactors and correlations for the validation of CFD codes. The test section is a 1 m long transparent pipe that is surrounded by an 18×18 cm rectangular “aquarium” filled with distilled water for refraction correction. High-speed camera (HSC) recording was used to gather data about condensing bubbles including bubble diameter, shape and rising velocity. Steam was injected via two different vertical injection nozzles with an inner diameter of 4 and 6 mm, respectively, directly into the center of the test section. The present experiments were carried out at three different steam superficial velocities, water superficial velocities and water temperatures leading to bubble diameters up to 50 mm and bubble relative velocities around 1 m/s. The measurements enabled the calculation of bubble Reynolds and Nusselt numbers and comparison with correlations given in literature. Even though significant differences could be observed between the two injection nozzles with respect to the bubble’s diameter and velocity, the Nusselt and Reynolds numbers are in the same range of values. The bigger bubbles of the 6 mm with respect to the 4 mm nozzle are almost neutralized by the lower rising velocities.

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