Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.

Author(s):  
James W. Morgan

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.


Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.


Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.


Radiocarbon ◽  
1995 ◽  
Vol 37 (2) ◽  
pp. 497-504 ◽  
Author(s):  
Mihály Veres ◽  
Ede Hertelendi ◽  
György Uchrin ◽  
Eszter Csaba ◽  
István Barnabás ◽  
...  

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.


Author(s):  
Young Ho Cho ◽  
Hyun-Seok Ko ◽  
Suk-Hoon Kim ◽  
Chang Sun Kang ◽  
Joo Hyun Moon ◽  
...  

The cost-effective reduction of occupational radiation dose (ORD) at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORD data of existing plants. Through the data analysis, it is required to identify what are the jobs of repetitive high ORD at the nuclear power plant. In this study, Percentile Rank Sum Method (PRSM) is proposed to identify repetitive high ORD jobs, which is based on non-parametric statistical theory. As a case study, the method is applied to ORD data of maintenance and repair jobs at Kori units 3 and 4 that are pressurized water reactors with 950 MWe capacity and have been operated since 1986 and 1987, respectively in Korea. The results was verified and validated, and PRSM has been demonstrated to be an efficient method of analyzing the data.


2021 ◽  
Author(s):  
Jin Feng Huang

Abstract After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR. The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR. We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity. The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor. In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.


Author(s):  
Robert Engel ◽  
Karl Zichanowicz

Since the 1970s power uprates have been employed to enhance the electricity output of nuclear power plants. Extended power uprate is defined as an increase of reactor thermal power in excess of 7% up to about 20% of the original plant licensed thermal power. Such power uprates generally require significant modification of major plant equipment and were initially implemented at boiling water reactor plants approximately 15 years ago. The early experience with extended power uprates was very positive. However, in recent years an increased number of failures of nuclear power plant components totally or partly caused by those power uprates have been reported. This paper presents the issues and operating experience with extended power uprate at the Leibstadt Boiling Water Reactor Nuclear Power Plant in Switzerland concerning the mechanical and electrical equipment. Failures such as fretting-wear and fatigue damage due to an increased vibration level, structural damage due to thermal overheating as well as shortened maintenance intervals due to elevated temperature and increased pump speed are reported. As a summary, the failures of mechanical and electrical components caused by extended power uprate experienced at the plant during the last years are related to plant reliability and have not adversely impacted plant safety.


2019 ◽  
Vol 21 (3) ◽  
pp. 119
Author(s):  
Alexandre Ezzidi Nakata ◽  
Masanori Naitoh ◽  
Chris Allison

Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation


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