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Nuclide composition non-uniformity in used nuclear fuel for considerations in pyroprocessing safeguards
The probability of Type-I error for the plutonium Material Unaccounted For (MUF) is evaluated by the hypothesis testing method as a function of the sizes of powder particles and granules, which are dominant parameters to determine the sample size. The results show the probability of Type-I error is occasionally greater than 5%. However, increasing granule sample sizes could surmount the weakness of material accounting because of the non-uniformity of nuclide composition.
7/4/2019 3:54:25 AM +00:00
Development of the nuclear safety trust indicator
This study went beyond making an indicator simply based on theoretical arguments, and explored a wide spectrum of different types of perceptions about energy safety to make a concept of energy safety for the Korean society. The energy safety schemata of people can be divided into three types. Type1 is concern about multi-level risks-responsibility-centric, type2 is concern about security and personal burden-expertise-centric, and type3 is concern about health and personal burden-responsibility-centric.
7/4/2019 3:54:06 AM +00:00
Experimental simulation of activity release from leaking fuel rods
The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leaking fuel rods under steady state and transient conditions in the spent fuel pool. The experimental rig included an electrically heated fuel rod with different defects and a cooling system. The fission product transport was simulated by potassium-chloride.
7/4/2019 3:53:47 AM +00:00
How should the regulatory defaults be set?
This paper, the extension of the previous work [1], focuses on the effects of different levels of conservatism implicated in regulatory defaults on the estimates of risk. According to the postulated behaviors of regulated parties and the diversity of interests of regulators, in particular, various measures for evaluating the effect of conservatism in defaults are developed and their properties are explored.
7/4/2019 3:53:30 AM +00:00
Radiotoxicity flux and concentration as complementary safety indicators for the safety assessment of a rock-cavern type LILW repository
This study presents a practical application of complementary safety indicators, which can be applied in a safety assessment of a radioactive waste repository by excluding a biosphere simulation and comparing the artificial radiation originating from the repository with the background natural radiation. Complementary safety indicators (radiotoxicity flux from geosphere and radiotoxicity concentration in seawater) were applied in the safety assessment of a rock-cavern type low and intermediate level radioactive waste (LILW) repository in the Republic of Korea.
7/4/2019 3:53:08 AM +00:00
Multi-unit Level 3 probabilistic safety assessment: Approaches and their application to a six-unit nuclear power plant site
The importance of performing Level 3 probabilistic safety assessments (PSA) along with a general interest in assessing multi-unit risk has been sharply increasing after the Fukushima Daiichi nuclear power plant (NPP) accident. However, relatively few studies on multi-unit Level 3 PSA have been performed to date, reflecting limited scenarios of multi-unit accidents with higher priority.
7/4/2019 3:52:51 AM +00:00
Three dimensional analysis of temperature effect on control rod worth in TRR
In this paper, three-dimensional neutronic calculations were performed in order to calculate the dependency of CRW on the temperature of fuel and moderator and the moderator void. Calculations were performed using the known MTR_PC computer codes in the core configuration 61 of TRR. The dependency of CRW on the fuel temperature in the range of 20e340 C and the moderator temperature of each control rods were studied.
7/4/2019 3:52:35 AM +00:00
ROSA/LSTF test and RELAP5 code analyses on PWR steam generator tube rupture accident with recovery actions
Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of highpressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondaryside pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondaryside pressures through sensitivity analyses with the RELAP5 code.
7/4/2019 3:52:15 AM +00:00
Decontamination of radioactive wastewater by two-staged chemical precipitation
This article presented two-staged chemical precipitation for radioactive wastewater decontamination by using chemical agents. The total amount of radioactive wastewater was 35 m3 , and main radionuclides were Cs-137, Cs-134, and Co-60. Initial radioactivity concentration of the liquid waste was 2264, 17, and 9 Bq/L for Cs-137, Cs-134 and Co-60, respectively. Potassium ferrocyanide, nickel nitrate, and ferrum nitrate were selected as chemical agents at high pH levels 8e10 according to the laboratory jar tests.
7/4/2019 3:52:00 AM +00:00
Monte Carlo burnup and its uncertainty propagation analyses for VERA depletion benchmarks by McCARD
In this paper, the quadratic extrapolation/quadratic interpolation (QEQI) method is proposed as a new high-order depletion scheme. In order to examine the effectiveness of the newly-implemented depletion modules in McCARD, four problems in the VERA depletion benchmarks are solved by CEBE/MEM, CEBE/CRAM, LEQI/ MEM, QEQI/MEM, and QDM for gadolinium isotopes. From the comparisons, it is shown that the QEQI/ MEM predicts kinf's most accurately among the test cases.
7/4/2019 3:51:43 AM +00:00
Precise prediction of radiation interaction position in plastic rod scintillators using a fast and simple technique: Artificial neural network
The mean absolute error was measured less than 2.5 and 5.5. The correlation coefficient was calculated 0.998 and 0.984, respectively. Also, the ANN technique was confirmed by leave-one-out (LOO) method with 1% error. These results presented the superiority of the ANN method in comparison with NLR and the other methods. The technique and set up used are simpler and faster than other the previous position sensitive detectors. Thus, the time, cost and shielding and electronics requirements are minimized and optimized.
7/4/2019 3:51:27 AM +00:00
Measurement of vibration and stress for APR-1400 reactor internals
This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype.
7/4/2019 3:51:10 AM +00:00
Nondestructive inspection of spent nuclear fuel storage canisters using shear horizontal guided waves
Nondestructive inspection (NDI) is an integral part of structural integrity analyses of dry storage casks that house spent nuclear fuel. One significant concern for the structural integrity is stress corrosion cracking in the heat-affected zone of welds in the stainless steel canister that confines the spent fuel. In situ NDI methodology for detection of stress corrosion cracking is investigated, where the inspection uses a delivery robot because of the presence of the harsh environment and geometric constrains inside the cask protecting the canister.
7/4/2019 3:50:48 AM +00:00
Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor
The main purpose of this work is to use the MTR_PC package to evaluate the effect of the partially insertion of the control rod on the neutronic parameters at the operating core of the Tehran Research Reactor. The simulation results show that by increasing the insertion of control rods (bank) in the core, the absolute values of power peaking factor, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time decreased.
7/4/2019 3:50:32 AM +00:00
An accident diagnosis algorithm using long short-term memory
This study proposes an algorithm for accident diagnosis using long short-term memory (LSTM), which is a kind of RNN, which improves the limitation for time reflection. The algorithm consists of preprocessing, the LSTM network, and postprocessing. In the LSTM-based algorithm, preprocessed input variables are calculated to output the accident diagnosis results. The outputs are also postprocessed using softmax to determine the ranking of accident diagnosis results with probabilities.
7/4/2019 3:50:11 AM +00:00
Planning of alternative countermeasures for a station blackout at a boiling water reactor using multilevel flow modeling
The second capability can be applied to find originally unassociated means to achieve a goal. This is vital in a situation where all designed means have failed. A technique of procedure synthesis can be used to express identified alternatives as a series of operations. A case of station blackout occurring at the boiling water reactor is described. An MFM model of a boiling water reactor is built according to the analysis of goals and functions. The accident situations are defined by the model, and several alternative countermeasures in terms of operating procedures are generated to achieve the goal of core cooling.
7/4/2019 3:49:51 AM +00:00
Development of an information reference system using reconstruction models of nuclear power plants
Three-dimensional (work-site) reconstruction models of dismantling fields are useful for workers to observe the conditions of dismantling field situations without visiting the actual fields. This study, based on AR and work-site reconstruction models, developed and evaluated an information reference system. The evaluation consists of questionnaires and interview surveys administered to six nuclear power plant workers who used this system, along with a scenario.
7/4/2019 3:49:29 AM +00:00
Criticality benchmark of McCARD Monte Carlo code for light-water-reactor fuel in transportation and storage packages
In this paper, McCARD code was verified using various models listed in the NUREG/CR-6361 benchmark guide, which provides specifications for single pin-cells, single assemblies, and the whole core classified depending on the nuclear properties and structural characteristics. McCARD code was verified by comparing its results with those of SCALE code for single pin-cell and single assembly benchmark problems.
7/4/2019 3:49:08 AM +00:00
Development of simulation-based testing environment for safety-critical software
In this study, a software testing method using a simulation-based software test bed is proposed. The test bed is developed by emulating the microprocessor architecture of the programmable logic controller used in NPP safety-critical applications and capturing its behavior at each machine instruction. The effectiveness of the proposed method is demonstrated via a case study. To represent the possible states of software input and the internal variables that contribute to generating a dedicated safety signal, the software test cases are developed in consideration of the digital characteristics of the target system and the plant dynamics.
7/4/2019 3:48:43 AM +00:00
Parameter identifiability of Boolean networks with application to fault diagnosis of nuclear plants
In this article, a sufficient condition for parameter identifiability of BN is first proposed, based on which the sufficient condition for fault identifiability of a sensor network is given. Then, the fault identifiability condition induces a sensor selection strategy for sensor selection. Finally, the theoretical result is applied to the fault diagnosiseoriented sensor selection for a nuclear heating reactor plant, and both the numerical computation and simulation results verify the feasibility of the newly built BN-based sensor selection strategy.
7/4/2019 3:48:30 AM +00:00
Towards defining a simplified procedure for COTS system-on-chip TID testing
In this paper we report the test setups, procedures and results for TID testing of a SoC microcontroller both using standard 60Co and low-energy protons beams. This paper specifically points out the differences in the test methodology and in the challenges between TID testing with proton beam and with the conventional gamma ray irradiation.
7/4/2019 3:48:08 AM +00:00
Monte Carlo simulations of criticality safety assessments of transuranic element storage in a pyroprocess facility
In this study, criticality safety assessments of the potential for storing transuranic element (TRU) ingots via a pyroprocess were evaluated to determine the appropriate TRU storage design parameters, in this case the ratio of the TRU ingot height to the radius and the number of TRU ingot canisters stacked within a container. Various accident situations were modeled over a modeling period of 5 years for a cumulative inventory of TRU ingots with various water densities in submerged containers and with various pitches between the containers in the facility.
7/4/2019 3:47:52 AM +00:00
Sensitivity studies in spent fuel pool criticality safety analysis for APR-1400 nuclear power plants
The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion.
7/4/2019 3:47:32 AM +00:00
Neutronic assessment of BDBA scenario at the end of Isfahan MNSR core life
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 (beff), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod.
7/4/2019 3:47:12 AM +00:00
Development of earthquake instrumentation for shutdown and restart criteria of the nuclear power plant using multivariable decision-making process
This article presents a new design of earthquake instrumentation that is suitable for quick decisionmaking after the seismic event at the nuclear power plant (NPP). The main objective of this work is to ensure more availability of the NPP by expediting walk-down period when the seismic wave is incident. In general, the decision-making to restart the NPP after the seismic event requires more than 1 month if an earthquake exceeds operating basis earthquake level.
7/4/2019 3:46:53 AM +00:00
A top-down iteration algorithm for Monte Carlo method for probability estimation of a fault tree with circular logic
The Monte Carlo method for fault tree quantification tends to take a long time because it repeats the calculation for a large number of samples. Herein, proposal is made to improve the quantification algorithm of a fault tree with circular logic. We developed a top-down iteration algorithm that combines the characteristics of the topdown approach and the iteration approach, thereby reducing the computation time of the Monte Carlo method.
7/4/2019 3:46:38 AM +00:00
Reactivity balance for a soluble boron-free small modular reactor
The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.
7/4/2019 3:46:23 AM +00:00
Development of an evaluation method for nuclear fuel debrisefiltering performance
This study presents an improved filtering test and an efficiency calculation method to fairly compare fuel-filtering efficiency regardless of the vendor's filtering features. To enhance the reliability of our evaluation, we established requirements for the test method and had a facility constructed according to the requirements.
7/4/2019 3:46:04 AM +00:00
Experimental study of bubble behaviors and CHF on printed circuit board (PCB) in saturated pool water at various inclination angles
The surface inclination angles were 0, 45, 90, 135, and 180. Results showed the Onset of Nucleate Boiling (ONB) heat flux increased with increasing heater orientation from 0 to 90, while early ONB occurred when the heater faced downwards (135 and 180). The nucleate boiling was observed to be unstable at low heat flux (1e21% of CHF) and changed into typical boiling when the heat flux was above 21% of CHF.
7/4/2019 3:45:45 AM +00:00
A study of neutron activation analysis compared to inductively coupled plasma atomic emission spectrometry for geological samples in Iran
In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM.
7/4/2019 3:45:30 AM +00:00