scholarly journals PREDICTION OF OPERATION LIFE EXTENSION OF HEAT POWER EQUIPMENT

2020 ◽  
Vol 2 (61) ◽  
pp. 51-60
Author(s):  
V. Skalozubov ◽  
◽  
D. Pirkovsky ◽  
M. Alali ◽  
R. Algerby ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Accumulation Rate ◽  
Reactor Vessel ◽  
Life Extension ◽  
Normal Operation ◽  
Cyclic Loads ◽  
Power Equipment ◽  
Heat Power ◽  
Operation Conditions ◽  
Accident Conditions

Based on the analysis of known researches, it is revealed that the quantity and accumulation rate of cyclic thermal and dynamic loads in the transient modes of normal operation conditions, when violating normal operation conditions and in accident conditions (except for the nuclear reactor vessel) are the key factors of prediction of operation life extension for a heat power equipment (heat exchangers, pumps, armature). The method for predictive estimation of terms of operation life extension of a heat power equipment depending on stress amplitudes in transient and accident conditions, quantity and accumulation rate of cyclic loads, strength metal parameters of a heat power equipment vessels (except for a reactor vessel) is provided. The method is implemented on the example of steam generators of WWERs and using operational data of South-Ukraine-1 (by 2010). Admissible accumulation rate of cyclic loads during operation life extension by 30, 40 and 50 years is a result. The results define insufficient substantiation of nuclear power plant operation in the “maneuverable” modes with a variable reactor power. In this case, the quantity of cyclic equipment loads increases dramatically, and terms of safe operation are limited. The developed method and the obtained results of prediction of operation life extension of heat power equipment can be used for industry programs to extend the operation of Ukrainian nuclear power plants, as well as to improve the regulatory documents governing the conditions and requirements for acceptable safe extension of the operation life of heat power equipment of nuclear and thermal power enterprises. Further improvement of the method proposed in the work for predicting operation life extension of heat power equipment can be based on the development of methods for analyzing the reliability of heat power equipment and databases on operation disturbances.The materials of the presented work are used in the educational process for the training, retraining and advanced training of specialists in the energy industry.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

scholarly journals Analysis of Loss of Offsite Power Events at China’s Nuclear Power Plants

2018 ◽  
Vol 10 (8) ◽  
pp. 2680 ◽  
Author(s):  
Feng Jiao ◽  
Shanshan Ding ◽  
Juan Li ◽  
Lixin Zheng ◽  
Qinghua Zhang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Electric Power System ◽  
Lessons Learned ◽  
Normal Operation ◽  
Emergency Power ◽  
Station Blackout ◽  
Accident Conditions

The function of the electric power system of nuclear power plants (NPPs) is to provide safe and reliable electricity for the equipment both in normal operation and accident conditions, and to provide emergency power for nuclear safety-related systems to maintain the safety of NPPs. Station blackout (SBO) occurs when loss of offsite power (LOOP) happens concurrently with unavailability of the onsite emergency alternating current (ac) power. LOOP is a precursor of SBO which rarely occurs but contributes significantly to reactor core damage frequency (CDF). Collecting and analyzing all LOOP events in NPPs of China from 1993 to 2017, this paper summarizes the common features of the LOOP events, and identifies the weaknesses and lessons learned from these events. Conclusions and experience feedback suggestions are put forward for improving the reliability of the offsite power supply of NPPs in China.

scholarly journals Improving Nuclear Power Plant Safety with FeCrAl Alloy Fuel Cladding

MRS Advances ◽  
2017 ◽  
Vol 2 (21-22) ◽  
pp. 1217-1224 ◽  
Author(s):  
Raul B. Rebak ◽  
Kurt A. Terrani ◽  
William P. Gassmann ◽  
John B. Williams ◽  
Kevin L. Ledford
Keyword(s):  
Nuclear Power ◽  
National Laboratory ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Oak Ridge ◽  
Plant Safety ◽  
Operation Conditions ◽  
Fecral Alloy ◽  
Nuclear Power Plant Safety

ABSTRACTThe US Department of Energy (DOE) is partnering with fuel vendors to develop enhanced accident tolerant nuclear fuels for Generation III water cooled reactors. In comparison with the standard current uranium dioxide and zirconium alloy system UO2-Zr), the proposed alternative accident tolerant fuel (ATF) should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory and their partners have proposed to replace zirconium based alloy cladding in current commercial power reactors with an iron-chromium-aluminum (FeCrAl) alloy cladding such as APMT. The use of FeCrAl alloys will greatly reduce the risk of operating the power reactors to produce electricity.

scholarly journals Hot particles analysis originating from failed and damaged fuels

2005 ◽  
Vol 3 (1) ◽  
pp. 106-117 ◽  
Author(s):  
Anikó Kerkápoly ◽  
Nóra Vajda ◽  
Tamás Pintér ◽  
Pintér Csordás
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Fission Products ◽  
Failure Detection ◽  
Normal Operation ◽  
Primary Coolant ◽  
Hot Particles ◽  
Operation Conditions ◽  
Transmutation Products ◽  
Primary Water

AbstractThe increase of activities of fission products and transmutation products in the primary coolant of a nuclear power plant indicates the presence of fuel rod failures. The measurement of the activity concentration of the primary coolant was able to detect fuel failures in the reactor core. Microanalytical methods for examining individual hot particles have been developed and applied to fuel failure detection under normal operation conditions as well as during the severe fuel damage that occurred in the cleaning tank incident at Unit 2 of NPP Paks in April 2003. Several faulty fuel rods can be detected simultaneously by the characterization of individual hot particles originating from the primary water. The analysis of particles originating from the damaged fuels provides information relating to the dissolution process of the fuel debris.

scholarly journals Estimation of Tritium Production in VVER-440 Reactor Core During Normal Operation

2019 ◽  
Vol 9 (2) ◽  
pp. 10-16
Author(s):  
Štefan Čerba ◽  
Jakub Lüley ◽  
Branislav Vrban ◽  
Filip Osuský ◽  
Vladimír Nečas
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Normal Operation ◽  
Boron Steel ◽  
Detailed Calculation ◽  
Primary Coolant ◽  
Tritium Production ◽  
Operation Conditions

Slovakia as one of the world leading countries in the share of nuclear power in electricityproduction and currently operates 2 nuclear power plants, each with 2 VVER-440 units. In addition to these reactors there are 2 VVER-440 units under construction and 2 units in decommissioning. The VVER-440 technology features thermal neutron spectrum, low enriched uranium dioxide fuel and light-water coolant, diluted boric acid and 37 emergency reactivity control assemblies with boron steel absorber. Due to the presence of 10B in the coolant/moderator which has high thermal neutron capture cross-section, the absorption of neutron on these atoms may lead to tritium production. Tritiumstrongly contributes to the level of radioactivity of the primary coolant, therefore the NPP staff must have appropriate knowledge of its production during operation. This paper focuses on the estimation of the tritium production for a specific scenario of the operation of the 3rd unit of Mochovce NPP. For simulations the SCALE6 system is used with the detailed calculation model developed at the B&J NUCLEAR ltd. company. The calculations presented in the paper are performed using self-shielded multi-group cross-section libraries, taking into account the operation conditions of Mochovce unit 3 NPP in the first fuel campaign.

scholarly journals ALTERNATIVE METHOD OF STRENGTH ANALYSIS FOR A HEAT-POWER EQUIPMENT UNDER CYCLIC LOADS

2020 ◽  
Vol 1 (60) ◽  
pp. 96-102
Author(s):  
V. Skalozubov ◽  
◽  
D. Pirkovsky ◽  
M. Alali ◽  
R. Algerby ◽  
...  
Keyword(s):  
Strength Analysis ◽  
Cyclic Loads ◽  
Power Equipment ◽  
Heat Power ◽  
Alternative Method

scholarly journals Transient Study on the HTR-PM with TINTE-vPower Coupling Code Package

2020 ◽  
Vol 2020 ◽  
pp. 1-14
Author(s):  
Jun Sun ◽  
Ximing Sun ◽  
Yanhua Zheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Flow ◽  
Nuclear Power ◽  
Data Exchange ◽  
Reactor Power ◽  
Normal Operation ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Operation Conditions

The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

Safety Analysis of Irradiated Nuclear Fuel Transportation Container

2007 ◽  
Author(s):  
E. Uspuras ◽  
S. Rimkevicius
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Analysis ◽  
Normal Operation ◽  
Spent Fuel ◽  
Safety Level ◽  
Fuel Assemblies ◽  
Accident Conditions ◽  
Transportation Container

Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit 1 on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level.

Characterization of Advanced Steels as Accident Tolerant Cladding for Light Water Reactor Nuclear Fuel

2015 ◽  
Author(s):  
Raul B. Rebak
Keyword(s):  
Zirconium Alloy ◽  
Fission Products ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Operation Conditions ◽  
Time Period ◽  
Environmental Test ◽  
Near Term ◽  
Accident Conditions

After the tsunami incident in Fukushima in March 2011, the international community is set to identify appropriate nuclear materials with increased accident tolerance with respect to the traditional UO2-zirconium alloy fuel system permitting loss of active cooling for a considerably longer time period, while maintaining or improving the fuel performance during normal operations. The researched safety characteristics of these advanced fuels are mainly: (a) Improved reaction kinetics with steam; (b) Slower hydrogen production rate; and (c) Enhanced retention of fission products. In the US the Department of Energy is supporting the development of an improved cladding using advanced steels such as the iron-chromium-aluminum (FeCrAl) alloy system. Environmental test results show that FeCrAl alloys are highly resistant to corrosion and environmental cracking under normal operation conditions and extremely resistant to attack by steam under accident conditions. That is, the replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant fuels.

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