14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

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.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

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.

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

2012 ◽  
Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Boiling Water ◽  
Pressurized Water ◽  
Water Reactor ◽  
Assessment Program ◽  
Water Reactors ◽  
The U.S ◽  
Reactor Internals

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.

Use of Percentile Rank Sum Method in Identifying Repetitive High Occupational Radiation Dose Jobs in a Nuclear Power Plant

2004 ◽  
Author(s):  
Young Ho Cho ◽  
Hyun-Seok Ko ◽  
Suk-Hoon Kim ◽  
Chang Sun Kang ◽  
Joo Hyun Moon ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Nuclear Power ◽  
Cost Effective ◽  
Pressurized Water ◽  
Percentile Rank ◽  
Water Reactors ◽  
The Cost ◽  
Occupational Radiation

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.

scholarly journals NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE

2019 ◽  
Vol 21 (3) ◽  
pp. 119
Author(s):  
Alexandre Ezzidi Nakata ◽  
Masanori Naitoh ◽  
Chris Allison
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Benchmark Problems ◽  
Joint Research ◽  
Pressurized Water ◽  
Water Reactors ◽  
Reactor Types ◽  
New Generation

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

Research on the Computerized Symptom-Oriented Procedures of Nuclear Power Plants

2008 ◽  
Author(s):  
Fei Liu ◽  
Zhijian Zhang ◽  
Minjun Peng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Results ◽  
Pressurized Water ◽  
Water Reactors ◽  
Steam Generator Tubes ◽  
Emergency Operating

New methods of information presentation and interface design are changing the working conditions in the modern Nuclear Power Plant (NPP) control room. Symptom-oriented EOPs (SOPs) with their structures and practical application are described. The Computerized Symptom-oriented Operating Procedures (CSOP) is researched, which can help the operator analyze all the symptom signals of steam generator tubes rupture (SGTR) and provide the computerized procedures corresponding to the symptom signals. This paper analyzes the accident of SGTR; the accident management of SGTR is important in reactor safety because SGTR is one of the relatively high-frequency events in pressurized water reactors PWRs. The symptom signals of SGTR and the possible accidents corresponding to the symptom signals are analyzed. The homologous measures of symptoms are summarized. The disposal of SGTR adopts the method of based on symptoms. The programs are developed by VxWorks that is a real-time operating system. The debugging of programs is processed on simulator. The test results indicated that the programs can provide operating procedures according to the symptoms of accidents. After adopting the Computerized Emergency Operating Procedures, the labor intensity and mental burden of operators are lightened. Computerized Emergency Operating Procedures can enhance the reliability, safety and efficiency of Nuclear Power Plant.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

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.

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