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Sensitivity analyses of the use of different neutron absorbers on the main safety core parameters in mtr type research reactor
In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, B4C, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations.
7/4/2019 6:22:06 AM +00:00
Semisupervised classification for fault diagnosis in nuclear power plants
In this paper, a fault diagnosis scheme based on semisupervised classification (SSC) scheme is developed. In this scheme, new measurements collected from the plant are integrated with data observed under fault conditions to train the SSC models. The trained models are subsequently applied to new measurements for fault diagnosis. In comparison with supervised classifiers, the proposed scheme requires significantly fewer data collected under fault conditions to train the classifier.
7/4/2019 6:21:49 AM +00:00
Seismic isolation of nuclear power plants
The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities.
7/4/2019 6:21:30 AM +00:00
Seismic isolation of lead cooled reactors: The european project siler
This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the nonisolated parts of the plant, such as the pipe expansion joints and the joint-cover of ư.
7/4/2019 6:21:07 AM +00:00
Second atlas domestic standard problem (DSP-02) for a code assessment
This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.
7/4/2019 6:20:44 AM +00:00
Scanning electron microscopy analysis of fuel matrix interaction layers
In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U–7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U–7Mo dispersion fuel elements with pure Al, Al–2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples.
7/4/2019 6:20:21 AM +00:00
Safety classification of systems, structures, and components for pool type research reactors
The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.
7/4/2019 6:20:07 AM +00:00
Safety aspects of intermediate heat transport and decay heat removal systems of sodium cooled fast reactors
Twenty sodium-cooled fast reactors (SFRs) have provided valuable experience in design, licensing, and operation. This paper summarizes the important safety criteria and safety guidelines of intermediate sodium systems, steam generators, decay heat removal systems and associated construction materials and in-service inspection. The safety criteria and guidelines provide a sufficient framework for design and licensing, in particular by new entrants in SFRs.
7/4/2019 6:19:53 AM +00:00
Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions
The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition.
7/4/2019 6:19:38 AM +00:00
Safety analysis methodology for aged candu nuclear reactors
This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models.
7/4/2019 6:19:23 AM +00:00
Round robin analyses on stress intensity factors of inner surface cracks in welded stainless steel pipes
The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.
7/4/2019 6:19:03 AM +00:00
Reliability data update using condition monitoring and prognostics in probabilistic safety assessment
This article provides enabling techniques to solidify a method to provide timeand condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs).
7/4/2019 6:18:48 AM +00:00
Reduction of radioactive waste from remediation of uranium contaminated soil
A high content of calcium in the waste solution was precipitated by adding sulfuric acid. The second waste can be significantly reduced by using sorption and desorption techniques on ampholyte resin S-950 prior to the precipitation of uranium at pH 3.0.
7/4/2019 6:18:35 AM +00:00
Recycling process of U3O8 powder in MnO-Al2O3 doped large grain UO2 pellets
The effect of various process variables on the powder properties of recycled U3O8 from MnO-Al2O3 doped large grain UO2 pellets and the effect of those recycled U3O8 powders on the sintered density and grain size of MnO-Al2O3 doped large grain UO2 pellets have been investigated. The evolution of morphology, size, and BET surface area of the recycled U3O8 powders according to the respective variation of the thermo-mechanical treatment variables of oxidation temperature, powder milling, and sequential cyclic heat treatment of oxidation and then reduction was examined.
7/4/2019 6:18:08 AM +00:00
Recent improvements in the cupid code for a multi dimensional two-phase flow analysis of nuclear reactor components
In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.
7/4/2019 6:17:51 AM +00:00
Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks
The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks.
7/4/2019 6:17:37 AM +00:00
Raim - A model for iodine behavior in containment under severe accident condition
This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases.
7/4/2019 6:17:15 AM +00:00
Radiological safety assessment of transporting radioactive wastes to the gyeongju disposal facility in Korea
The important part of the radiological safety assessment is related to the exposure dose assessment for the incidentfree (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public.
7/4/2019 6:17:01 AM +00:00
Radioactive source security: Why do we not yet have a global protection system?
Security of radioactive sources has been an issue since the earliest days of safety regulation of such materials. Since the events of September 11 2001, some governments and regulatory bodies have been much more focussed on these issues and have introduced extensive and enhanced security arrangements. International organisations like the IAEA and WINS have worked hard to help States in this regard. However, only a minority of States have implemented statutory security systems for radioactive source security.
7/4/2019 6:16:48 AM +00:00
Quantitative observation of co current stratified two-phase flow in a horizontal rectangular channel
The main objective of this study is to investigate experimentally the two-phase flow characteristics in terms of the direct contact condensation of a steamewater stratified flow in a horizontal rectangular channel.
7/4/2019 6:16:31 AM +00:00
Px - an innovative safety concept for an unmanned reactor
The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents.
7/4/2019 6:16:12 AM +00:00
PWSCC growth assessment model considering stress triaxiality factor for primary alloy 600 components
By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.
7/4/2019 6:15:58 AM +00:00
Pump design and computational fluid dynamic analysis for high temperature sulfuric acid transfer system
In this study, we proposed a newly designed sulfuric acid transfer system for the sulfur-iodine (SI) thermochemical cycle. The proposed sulfuric acid transfer system was evaluated using a computational fluid dynamics (CFD) analysis for investigating thermodynamic/hydrodynamic characteristics and material properties. This analysis was conducted to obtain reliable continuous operation parameters; in particular, a thermal analysis was performed on the bellows box and bellows at amplitudes and various frequencies (0.1, 0.5, and 1.0 Hz).
7/4/2019 6:15:40 AM +00:00
Propagation of nuclear data uncertainties for pwr core analysis
An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for 235,238U, 239Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.
7/4/2019 6:15:24 AM +00:00
Prolongation of the bor 60 reactor operation
The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor
7/4/2019 6:15:10 AM +00:00
Probabilistic seismic assessment of base isolated npps subjected to strong ground motions of tohoku earthquake
The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.
7/4/2019 6:14:50 AM +00:00
Preparation and properties of the fast curing g-ray shielding materials based on polyurethane
In this study, fast-curing shielding materials were prepared with a two-component polyurethane matrix and a filler material of PbO through a one-step, laboratory-scale method. With an increase in the filler content, viscosity increased. However, the two components showed a small difference. Curing time decreased as the filler content increased.
7/4/2019 6:14:34 AM +00:00
Preliminary safety study of engineering scale pyroprocess facility
In this paper, the concept of the AFC facility was introduced, and its safety evaluations were performed. For the safety evaluations, anticipated accident events were selected, and environmental safety analyses were conducted for the safety of the public and workers. In addition, basic radiation shielding safety analyses and criticality safety analyses were conducted.
7/4/2019 6:14:19 AM +00:00
Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstances of NPPS
The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed.
7/4/2019 6:14:05 AM +00:00
Prediction of hydrogen concentration in containment during severe accidents using fuzzy neural network
The aim of this paper is to predict the hydrogen buildup within containment resulting from severe accidents. The prediction was based on NPPs of an optimized power reactor 1,000. The increase in the hydrogen concentration in severe accidents is one of the major factors that threaten the integrity of the containment.
7/4/2019 6:13:50 AM +00:00