• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1321-1329
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• 2016

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.276-286
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• 1999

A computer code PARS2 is developed for the analysis of PSDRS, which is the safety grade RHRS of HAMMER, and applied to the investigation of the relation between design parameters and performance of PSDRS. The concept of the heat transfer resistance network is applied in assessing the importance of the various heat transfer modes. From the analysis results, the qualitative relations between the PSDRS performance and design parameters are found and guidelines for the PSDRS design procedures are also proposed.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.145-150
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• 1996

Acoustic Emission (AE) technique was applied to stress corrosion cracking of Inconel 600 to investigate the AE capability of detecting crack growth and to obtain the relation between AE characteristics and crack mechanism. The specimens were heat-treated in two conditions (600$^{\circ}C$ for 30 hrs or 700 $^{\circ}C$ for 1 hr) and undergone CERT at two extension rates ( 2.5${\times}$10$^{-5}$ or 1.25${\times}$10$^{－4}$(mm/s)). It was found that the AE peak amplitude from plastic deformation was generally smaller than about 48dB (0.25mV), while Intergranular stress corrosion cracking (IGSCC) and ductile fracture produced higher values of 49 to 70dB (0.3mV to 3mV). The slopes of cumulative amplitude distribution (b-values) were linearly dependent on IGSCC susceptibility and the higher the susceptibility, the smaller the b-value. The monitoring of combined AE parameters such as event rate, amplitude, count and energy can provide effective means to clearly identify the transition from crack initiation and small crack growth to rapid growth of dominant cracks.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.55-62
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• 2014

A method for estimating software reliability for nuclear safety software is proposed in this paper. This method is based on the software reliability growth model (SRGM), where the behavior of software failure is assumed to follow a non-homogeneous Poisson process. Two types of modeling schemes based on a particular underlying method are proposed in order to more precisely estimate and predict the number of software defects based on very rare software failure data. The Bayesian statistical inference is employed to estimate the model parameters by incorporating software test cases as a covariate into the model. It was identified that these models are capable of reasonably estimating the remaining number of software defects which directly affects the reactor trip functions. The software reliability might be estimated from these modeling equations, and one approach of obtaining software reliability value is proposed in this paper.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2009-2016
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• 2020

A nuclear power plant is a large complex system with tens of thousands of components. To ensure plant safety, the early and accurate diagnosis of abnormal situations is an important factor. To prevent misdiagnosis, operating procedures provide the anticipated symptoms of abnormal situations. While the more severe emergency situations total less than ten cases and can be diagnosed by dozens of key plant parameters, abnormal situations on the other hand include hundreds of cases and a multitude of parameters that should be considered for diagnosis. The tasks required of operators to select the appropriate operating procedure by monitoring large amounts of information within a limited amount of time can burden operators. This paper aims to develop a system that can, in a short time and with high accuracy, select the appropriate operating procedure and sub-procedure in an abnormal situation. Correspondingly, the proposed model has two levels of prediction to determine the procedure level and the detailed cause of an event. Simulations were conducted to evaluate the developed model, with results demonstrating high levels of performance. The model is expected to reduce the workload of operators in abnormal situations by providing the appropriate procedure to ultimately improve plant safety.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.106-115
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• 1984

Core safety limits define reactor operating conditions and parameters that will assure fuel rod and reactor system's integrity. Limiting safety system settings (LSSS) programmed into reactor protection system (RPS) then ensure a rapid reactor trip to prevent or suppress conditions which might violate the core safety limits. Generation of the LSSS must properly take into account uncertainties in both calculated and measured parameters in order to assure, with an appropriate degree of confidence, that the RPS will protect the core safety limits. Reviewed in this report are Westinghouse RPS setpoint generation philosophy, methodology of safety limit development and LSSS generation procedure. The Westinghouse RPS trip setpoint generation methodology has been established based on the calculation of core safety limits and the selection of LSSS allowing appropriate uncertainties in a conservative manner. Such conservative values of setpoint assure a high degree of core protection against fuel melting and occurrence of DNB.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.109-118
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• 2023

In this paper, the evaluation method of fluidelastic instability (FEI) of newly designed steam generator tubing in pressurized water reactor (PWR) nuclear power plants is discussed. To obtain the parameters for prediction of the critical velocity of FEI for steam generator tubes, experimental research is carried out, and the design parameters are determined. Using CFD numerical simulation, the tube array scale of the model experiment is determined, and the experimental device is designed. In this paper, 7 groups of experiments with void fractions of 0% (water), 10%, 20%, 50%, 75%, 85% and 95% were carried out. The critical damping ration, fundamental frequency and critical velocity of FEI of tubes in flowing water were measured. Through calculation, the total mass and instability constant of the immersed tube are obtained. The critical damping ration measured in the experiment mainly included two-phase damping and viscous damping, which changed with the change in void fraction from 1.56% to 4.34%. This value can be used in the steam generator design described in this paper and is conservative. By introducing the multiplier of frequency and square root of total mass per unit length, it is found that the difference between the experimental results and the calculated results is less than 1%, which proves the rationality and feasibility of the calculation method of frequency and total mass per unit length in engineering design. Through calculation, the instability constant is greater than 4 when the void fraction is less than 75%, less than 4 when the void fraction exceeds 75% and only 3.04 when the void fraction is 95%.