scholarly journals Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor coolant system and systems decontamination. Summary status report. Volume 1

1986 ◽  
Author(s):  
D.H. Doerge ◽  
R.L. Miller ◽  
K.S. Scotti
Keyword(s):  
Status Report ◽  
Nuclear Facility ◽  
Reactor Coolant ◽  
Coolant System

scholarly journals Decontamination of a Contaminated RCP Shaft Using the SP-HyBRID Process

Processes ◽  
2021 ◽  
Vol 9 (10) ◽  
pp. 1725
Author(s):  
Hee-Chul Eun ◽  
Na-On Chang ◽  
Wang-Kyu Choi ◽  
Sang-Yoon Park ◽  
Seon-Byeong Kim ◽  
...  
Keyword(s):  
Decontamination Factor ◽  
Hybrid Process ◽  
Waste Generation ◽  
Nuclear Facility ◽  
Experimental Conditions ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Pump Shaft ◽  
Shaft Surface ◽  
Coolant System

It is very important to minimize the waste generation for decontamination of the reactor coolant system in a nuclear facility. As an alternative to commercial decontamination technologies, an inorganic acid chemical decontamination (SP-HyBRID) process can be effectively applied to the decontamination because it can significantly reduce the waste generation. In this study, the decontamination of a contaminated reactor coolant pump shaft from a nuclear facility was conducted using the SP-HyBRID process. First, equipment for a mock-up test of the decontamination was prepared. Detailed experimental conditions for the decontamination were determined through the mock-up test. Under the detailed conditions, the contaminated shaft was successfully decontaminated. The dose rate on the shaft surface was greatly reduced from 1400 to 0.9 μSv/h, and the decontamination factor showed a very high value (>1500).

scholarly journals FATIGUE LIFE ASSESSMENT OF REACTOR COOLANT SYSTEM COMPONENTS BY USING TRANSFER FUNCTIONS OF INTEGRATED FE MODEL

2010 ◽  
Vol 42 (5) ◽  
pp. 590-599 ◽  
Author(s):  
Shin-Beom Choi ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
Myung-Jo Jhung ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Transfer Functions ◽  
Life Assessment ◽  
Fe Model ◽  
Reactor Coolant ◽  
Fatigue Life Assessment ◽  
System Components ◽  
Coolant System

Experimental study on transport behavior of cesium iodide in the reactor coolant system under LWR severe accident conditions

2020 ◽  
Vol 57 (12) ◽  
pp. 1287-1296
Author(s):  
Naoya Miyahara ◽  
Shuhei Miwa ◽  
Mélany Gouëllo ◽  
Junpei Imoto ◽  
Naoki Horiguchi ◽  
...  
Keyword(s):  
Experimental Study ◽  
Cesium Iodide ◽  
Severe Accident ◽  
Reactor Coolant ◽  
Transport Behavior ◽  
Accident Conditions ◽  
Coolant System

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

Sensitivity Analysis for Dynamical Response of Reactor Coolant System Based on OPTIMUS

2020 ◽  
Author(s):  
Yuan Yanli ◽  
Ye Xianhui ◽  
Li Lijuan ◽  
Yuan Feng
Keyword(s):  
Sensitivity Analysis ◽  
Steam Generator ◽  
Nuclear Power ◽  
Reactor System ◽  
Response Analysis ◽  
Dynamical Response ◽  
Calculation Error ◽  
Reactor Coolant ◽  
Input Parameters ◽  
Coolant System

Abstract The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis.

“Begging the Question” in Licensing Basis Accident Analyses

2018 ◽  
Author(s):  
Samuel Miranda
Keyword(s):  
Simultaneous Occurrence ◽  
Begging The Question ◽  
Reactor Coolant ◽  
The Third ◽  
American Nuclear Society ◽  
Safety Limit ◽  
Logical Fallacy ◽  
Coolant System

“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill. This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that: (1) reactor coolant system (RCS) pressure will not exceed its safety limit, and (2) no fuel damage will be incurred, and (3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault. The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips. “Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.

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