Science and Technology of Nuclear Installations https://www.hindawi.com The latest articles from Hindawi © 2017 , Hindawi Limited . All rights reserved. HTR-10GT Dual Bypass Valve Control Features and Decoupling Strategy for Power Regulation Mon, 13 Feb 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/9404636/ HTR-10GT is the development of HTR-10 reactor, which PCU will be a closed Brayton cycle with two-stage compression and heat recuperation. Bypass control method is adopted for rapid power regulation and safety protection. But quick opening of single bypass valve would inevitably lead to temperature shocks in multiple components especially at the reactor inlet and the recuperator core. Based on the regulating characteristics of each possible bypass valve, a dual bypass valve control scheme was proposed along with MIMO decoupling controller designed with diagonal matrix method. The system was modeled with Modelica; the DASSL code was used to solve the Differential and Algebraic Equations during simulations. System’s control characteristic was analyzed with classical linear control theory and theory applied on linearized system model. Further numerical simulations showed that cooperative functioning of two bypass valves could effectively limit the temperature variation during power regulation, while the decoupler could improve the control effect and the stability of the system. The results will be helpful for the future design of the control system of HTR-10GT or other closed Brayton cycle of the same kind. Xiao Li, Xiaoyong Yang, Youjie Zhang, and Jie Wang Copyright © 2017 Xiao Li et al. All rights reserved. RELAP5 Simulation of PKL Facility Experiments under Midloop Conditions Thu, 09 Feb 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/6140323/ Nuclear power plant risk has to be quantified in full power and in other modes of operation. This latter situation corresponds to low power and shutdown modes of operation in which the residual heat removal (RHR) system is required to extract the heat generated in the core. These accidental sequences are great contributors to the total plant risk. Thus, it is important to analyze the plant behavior to establish the accident mitigation measures required. In this way, PKL facility experimental series were undertaken to analyze the plant behavior in other modes of operation when the RHR is lost. In these experiments, the plant configurations were changed to analyze the influence of steam generators secondary side configurations, the temperature inside the pressurizer, and the inventory level on the plant behavior. Moreover, different accident management measures were proposed in each experiment to reach the conditions to restart the RHR. To understand the physical phenomena that takes place inside the reactor, the experiments are simulated with thermal-hydraulic codes, and this makes it possible to analyze the code capabilities to predict the plant behavior. This work presents the simulation results of four experiments included in PKL experimental series obtained using RELAP5/Mod3.3. J. F. Villanueva, S. Carlos, F. Sanchez-Saez, I. Martón, and S. Martorell Copyright © 2017 J. F. Villanueva et al. All rights reserved. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems Thu, 26 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/7275346/ A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD) uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA) is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results. Thomas Frosio, Thomas Bonaccorsi, and Patrick Blaise Copyright © 2017 Thomas Frosio et al. All rights reserved. Unbalance Compensation of a Full Scale Test Rig Designed for HTR-10GT: A Frequency-Domain Approach Based on Iterative Learning Control Thu, 26 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/3126738/ Unbalance vibrations are crucial problems in heavy rotational machinery, especially for the systems with high operation speed, like turbine machinery. For the program of 10 MW High Temperature gas-cooled Reactor with direct Gas-Turbine cycle (HTR-10GT), the rated operation speed of the turbine system is 15000 RPM which is beyond the second bending frequency. In that case, even a small residual mass will lead to large unbalance vibrations. Thus, it is of great significance to study balancing methods for the system. As the turbine rotor is designed to be suspended by active magnetic bearings (AMBs), unbalance compensation could be achieved by adequate control strategies. In the paper, unbalance compensation for the Multi-Input and Multi-Output (MIMO) active magnetic bearing (AMB) system using frequency-domain iterative learning control (ILC) is analyzed. Based on the analysis, an ILC controller for unbalance compensation of the full scale test rig, which is designed for the rotor and AMBs in HTR-10GT, is designed. Simulation results are reported which show the efficiency of the ILC controller for attenuating the unbalance vibration of the full scale test rig. This research can offer valuable design criterion for unbalance compensation of the turbine machinery in HTR-10GT. Ying He, Lei Shi, Zhengang Shi, and Zhe Sun Copyright © 2017 Ying He et al. All rights reserved. Challenge Analysis and Schemes Design for the CFD Simulation of PWR Tue, 24 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/5695809/ CFD simulation for a PWR is an important part for the development of Numerical Virtual Reactor (NVR) in Harbin Engineering University of China. CFD simulation can provide the detailed information of the flow and heat transfer process in a PWR. However, a large number of narrow flow channels with numerous complex structures (mixing vanes, dimples, springs, etc.) are located in a typical PWR. To obtain a better CFD simulation, the challenges created by these structural features were analyzed and some quantitative regularity and estimation were given in this paper. It was found that both computing resources and time are in great need for the CFD simulation of a whole reactor. These challenges have to be resolved, so two schemes were designed to assist/realize the reduction of the simulation burden on resources and time. One scheme is used to predict the combined efficiency of the simulation conditions (configuration of computing resources and application of simulation schemes), so it can assist the better choice/decision of the combination of the simulation conditions. The other scheme is based on the suitable simplification and modification, and it can directly reduce great computing burden. Guangliang Chen, Zhijian Zhang, Zhaofei Tian, Lei Li, and Xiaomeng Dong Copyright © 2017 Guangliang Chen et al. All rights reserved. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl Wed, 18 Jan 2017 14:29:13 +0000 http://www.hindawi.com/journals/stni/2017/3146985/ Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view. Shengli Chen and Cenxi Yuan Copyright © 2017 Shengli Chen and Cenxi Yuan. All rights reserved. Evaluation of ACPs in China Fusion Engineering Test Reactor Using CATE 2.1 Code Mon, 09 Jan 2017 12:57:57 +0000 http://www.hindawi.com/journals/stni/2017/2936069/ Activated corrosion products (ACPs) are the dominant radiation hazard in water-cooled fusion reactor under normal operation conditions and directly determine the occupational radiation exposure during operation and maintenance. Recently, the preliminary design of China Fusion Engineering Test Reactor (CFETR) has been just completed. Evaluation of ACPs is an important work for the safety of CFETR. In this paper, the ACPs analysis code CATE 2.1 was used to simulate the spatial distribution of ACPs along the blanket cooling loop of CFETR, in which the influence of adopting different pulse handling methods was researched. At last, the dose rate caused by ACPs around the blanket cooling loop was calculated using the point kernel code ARShield. The results showed that the dose rate under normal operation for 1.2 years at contact is 1.02 mSv/h and at 1 m away from pipe is 0.45 mSv/h. And after shutting down the reactor, there will be a rapid decrease of dose rate, because of the rapid decay of short-lived ACPs. Lu Li, Jingyu Zhang, Qingyang Guo, Xiaokang Zhang, Songlin Liu, and Yixue Chen Copyright © 2017 Lu Li et al. All rights reserved. Separation of Transformers for Class 1E Systems in Nuclear Power Plants Mon, 02 Jan 2017 13:00:39 +0000 http://www.hindawi.com/journals/stni/2017/3976049/ In order to supply electric power to the safety related loads, safety and reliability of onsite power have to be ensured for the safety function performance in nuclear power plants. Even though the existing electric power system of APR1400 meets the requirements of codes regarding Class 1E system, there is a room for improvement in the design margin against the voltage drop and short circuit current. This paper discusses the amount that the voltage drop and short circuit current occur in the existing electric power system of APR1400. Additionally, this paper studies with regard to the improved model that has the extra margin against the high voltage drop and short circuit current by separation of unit auxiliary transformer (UAT) and standby auxiliary transformer (SAT) for the Class 1E loads. The improved model of the electric power system by separation of UAT and SAT has been suggested through this paper. Additionally, effects of reliability and cost caused by the electric power system modification are considered. Sang-Hyun Lee and Choong-Koo Chang Copyright © 2017 Sang-Hyun Lee and Choong-Koo Chang. All rights reserved. Mechanical Properties in Nuclear Installation and the Relevant Measurement Methods Sun, 25 Dec 2016 12:20:04 +0000 http://www.hindawi.com/journals/stni/2016/1948507/ Yan Yang, Alejandro Clausse, Leon Cizelj, Xing Chen, and Parashuram Sahoo Copyright © 2016 Yan Yang et al. All rights reserved. Using CFD as Preventative Maintenance Tool for the Cold Neutron Source Thermosiphon System Mon, 19 Dec 2016 13:55:52 +0000 http://www.hindawi.com/journals/stni/2016/5452085/ The cold neutron source (CNS) system of the Open Pool Australian Light-Water (OPAL) reactor is a 20 L cryogenically cooled liquid deuterium thermosiphon system. The CNS is cooled by forced convective helium which is held at room temperature during stand-by (SO) mode and at ~20 K during normal operation (NO) mode. When helium cooling stops, the reactor is shut down to prevent the moderator cell from overheating. This computational fluid dynamics (CFD) study aims to determine whether the combined effects of conduction and natural convection would provide sufficient cooling for the moderator cell under the influence of reactor decay heat after reactor shutdown. To achieve this, two commercial CFD software packages using an identical CFD mesh were first assessed against an experimental heat transfer test of the CNS. It was found that both numerical models were valid within the bounds of experimental uncertainty. After this, one CFD model was used to simulate the thermosiphon transient condition after the reactor is shut down. Two independent shutdown conditions of different decay-heat power profiles were simulated. It was found that the natural convection and conduction cooling in the thermosiphon were sufficient for dissipating both decay-heat profiles, with the moderator cell remaining below the safe temperature of 200°C. Mark Ho, Yeongshin Jeong, Haneol Park, Guan Heng Yeoh, and Weijian Lu Copyright © 2016 Mark Ho et al. All rights reserved. Sizing of the Vacuum Vessel Pressure Suppression System of a Fusion Reactor Based on a Water-Cooled Blanket, for the Purpose of the Preconceptual Design Mon, 19 Dec 2016 11:11:26 +0000 http://www.hindawi.com/journals/stni/2016/8719695/ A methodology to preliminarily evaluate the size of the suppression tank and the relief pipes for a Vacuum Vessel Pressure Suppression System, to be adopted in a fusion reactor based on a water cooled blanket, is presented. The volume of the ST depends on the total energy of the water cooling system and it can be sized based on a required final pressure at equilibrium, by a simple energy balance. The pressure peak in the VV depends mainly on break area and the flow area of the relief pipes and some suggestions about the method for a preliminarily evaluation of their size are discussed. The computer code CONSEN has been used to perform a parametric study and to verify the methodology. Gianfranco Caruso and Fabio Giannetti Copyright © 2016 Gianfranco Caruso and Fabio Giannetti. All rights reserved. Analysis of Density Wave Oscillations in Helically Coiled Tube Once-Through Steam Generator Thu, 15 Dec 2016 14:26:31 +0000 http://www.hindawi.com/journals/stni/2016/3071686/ Helically coiled tube Once-Through Steam Generator (H-OTSG) is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of low and medium pressure. This paper presents research on density wave oscillations (DWO) in a typical countercurrent H-OTSG. Based on the steady-state calculation, the mathematical model of single-channel system was established, and the transfer function was derived. Using Nyquist stability criterion of the single variable, the stability cases were studied with an in-house computer program. According to the analyses, the impact law of the geometrical parameters to the system stability was obtained. RELAP5/MOD3.2 code was also used to simulate DWO in H-OTSG. The theoretical analyses of the in-house program were compared to the simulation results of RELAP5. A correction factor was introduced to reduce the error of RELAP5 when modeling helical geometry. The comparison results agreed well which showed that the correction is effective. Junwei Hao, Yaoli Zhang, Jianxiang Zheng, Zhiwei Zhou, Xuanyu Sheng, Gang Hong, Kai Ye, and Ning Li Copyright © 2016 Junwei Hao et al. All rights reserved. Cost Estimation and Efficiency Analysis of Korean CANDU Spent Fuel Disposal Alternatives in Consideration of Future Price Volatility Thu, 15 Dec 2016 13:21:29 +0000 http://www.hindawi.com/journals/stni/2016/3967572/ In Korea, spent fuel is temporarily stored in spent fuel pools at nuclear reactor sites and it is predicted to become saturated between 2020 and 2024. For this reason, four disposal alternatives (KRS-1, A-KRS-1, A-KRS-21, and A-KRS-22) have been developed in order to carry out the direct disposal of the CANDU spent fuel. The objective of this study is to conduct cost efficiency analysis of the disposal alternatives in consideration of price volatility for the radioactive waste repository. To derive future price volatility, this study used the ARIMA model. As a result, A-KRS-1 is the most efficient in terms of price per bundle using 2015 price. As for the results using ARIMA model, except in the case of KRS-1, the cost per bundle of A-KRS-1, A-KRS-21, and A-KRS-22 is decreased. Cost estimation using ARIMA model shows little change or decreases in cost while cost estimation using inflation rates for 2020 resulted in approximately 7.2% increases compared to 2015 for all options. As for the results of scenario analysis, A-KRS-1 earned 8,160 points, while A-KRS-22 followed closely behind with 7,980 points among the total 24,300 points. The results of this study provide invaluable policy data for any nation considering the construction of spent nuclear fuel repository. Sungsig Bang, Yanghon Chung, Dongphil Chun, Chulhong Kwon, and Sungjun Hong Copyright © 2016 Sungsig Bang et al. All rights reserved. Development of End Plug Welding Technique for SFR Fuel Rod Fabrication Thu, 15 Dec 2016 10:10:55 +0000 http://www.hindawi.com/journals/stni/2016/9549805/ In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters were developed for the end plug welding of SFR metallic fuel rods. A gas tungsten arc welding (GTAW) technique was adopted and the welding joint design was developed. In addition, the optimal welding conditions and parameters were established. Based on the establishment of the welding conditions, the GTAW technique was qualified for the end plug welding of SFR metallic fuel rods. Jung Won Lee, Jong Hwan Kim, Ki Hwan Kim, Jeong Yong Park, and Sung Ho Kim Copyright © 2016 Jung Won Lee et al. All rights reserved. Vibration Control of Nuclear Power Plant Piping System Using Stockbridge Damper under Earthquakes Tue, 13 Dec 2016 13:14:08 +0000 http://www.hindawi.com/journals/stni/2016/5014093/ Generally the piping system of a nuclear power plant (NPP) has to be designed for normal loads such as dead weight, internal pressure, temperature, and accidental loads such as earthquake. In the proposed paper, effect of Stockbridge damper to mitigate the response of piping system of NPP subjected to earthquake is studied. Finite element analysis of piping system with and without Stockbridge damper using commercial software SAP2000 is performed. Vertical and horizontal components of earthquakes such as El Centro, California, and Northridge are used in the piping analysis. A sine sweep wave is also used to investigate the control effects on the piping system under wide frequency range. It is found that the proposed Stockbridge damper can reduce the seismic response of piping system subjected to earthquake loading. Seongkyu Chang, Weipeng Sun, Sung Gook Cho, and Dookie Kim Copyright © 2016 Seongkyu Chang et al. All rights reserved. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology Wed, 07 Dec 2016 06:06:03 +0000 http://www.hindawi.com/journals/stni/2016/7328131/ The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS) were demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR) parameters is based on deterministic transport theory ( method) providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between -MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources. Mario Matijević, Dubravko Pevec, and Krešimir Trontl Copyright © 2016 Mario Matijević et al. All rights reserved. Progress of Particle Flow, Fluid/Solid Mechanics, and Heat Transfer in Advanced Gas/Water Nuclear Reactors Sun, 27 Nov 2016 13:01:53 +0000 http://www.hindawi.com/journals/stni/2016/2512634/ Nan Gui, Xiangdong Li, and Xin Tu Copyright © 2016 Nan Gui et al. All rights reserved. Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) Mon, 21 Nov 2016 06:05:43 +0000 http://www.hindawi.com/journals/stni/2016/5967831/ Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR). To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10) in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology. Chengxiang Guo, Liqiang Wei, Hong Li, Jianzhu Cao, and Sheng Fang Copyright © 2016 Chengxiang Guo et al. All rights reserved. Development of Vacuum Vessel Design and Analysis Module for CFETR Integration Design Platform Thu, 10 Nov 2016 13:58:44 +0000 http://www.hindawi.com/journals/stni/2016/5321057/ An integration design platform is under development for the design of the China Fusion Engineering Test Reactor (CFETR). It mainly includes the integration physical design platform and the integration engineering design platform. The integration engineering design platform aims at performing detailed engineering design for each tokamak component (e.g., breeding blanket, divertor, and vacuum vessel). The vacuum vessel design and analysis module is a part of the integration engineering design platform. The main idea of this module is to integrate the popular CAD/CAE software to form a consistent development environment. Specifically, the software OPTIMUS provides the approach to integrate the CAD/CAE software such as CATIA and ANSYS and form a design/analysis workflow for the vacuum vessel module. This design/analysis workflow could automate the process of modeling and finite element (FE) analysis for vacuum vessel. Functions such as sensitivity analysis and optimization of geometric parameters have been provided based on the design/analysis workflow. In addition, data from the model and FE analysis could be easily exchanged among different modules by providing a unifying data structure to maintain the consistency of the global design. This paper describes the strategy and methodology of the workflow in the vacuum vessel module. An example is given as a test of the workflow and functions of the vacuum vessel module. The results indicate that the module is a feasible framework for future application. Chen Zhu, Minyou Ye, Xufeng Liu, Shenji Wang, Shifeng Mao, Zhongwei Wang, and Yi Yu Copyright © 2016 Chen Zhu et al. All rights reserved. Radioactive Source Specification of Bushehr’s VVER-1000 Spent Fuels Thu, 10 Nov 2016 13:45:02 +0000 http://www.hindawi.com/journals/stni/2016/4579738/ Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressurized water reactor, was determined. For the depletion and decay calculations, ORIGEN code was utilized. The results are presented for burnups of 30 to 49 GWd/MTHM and different cooling times up to 100 years. According to these results, total activity of a spent fuel assembly with initial enrichment of 3.92%, burnup of 49 GWd/MTHM, and cooling time of 3 years is 1.92 × 1016 Bq. The results can be utilized specifically in transportation/storage cask design for spent fuel management of Bushehr nuclear power plant. Mahdi Rezaeian and Jamshid Kamali Copyright © 2016 Mahdi Rezaeian and Jamshid Kamali. All rights reserved. Thermal Analysis for the Dense Granular Target of CIADS Mon, 07 Nov 2016 13:21:03 +0000 http://www.hindawi.com/journals/stni/2016/5158610/ For the China Initiative Accelerator Driven System (CIADS), the energy of the protons is 250 MeV, and the current intensity will reach 10 milliamperes. A new concept of a dense granular spallation target is proposed for which the tungsten granules are chosen as the target material. After being bombarded with the accelerated protons from the accelerator, the tungsten granules with high-temperature flow out of the subcritical reactor and the heat is removed by the heat exchanger. One key issue of the target is to remove the 2.5 MW heat deposition safely. Another one is the heat exchange between the target and the subcritical reactor. Based on the model of effective thermal conductivity, a new thermal code is developed in Matlab. The new code is used to calculate the temperature field of the target area near active zone and it is partly verified by commercial CFD code Fluent. The result shows that the peak temperature of the target zone is nearly 740°C and the reactor and the target are proved to be uncoupled in thermal process. Kang Chen, Yongwei Yang, and Yucui Gao Copyright © 2016 Kang Chen et al. All rights reserved. The Investigation of Nonavailability of Passive Safety Systems Effects on Small Break LOCA Sequence in AP1000 Using RELAP5 MOD 4.0 Sun, 06 Nov 2016 08:39:44 +0000 http://www.hindawi.com/journals/stni/2016/7690151/ The passive safety systems of AP1000 are designed to operate automatically at desired set-points. However, the unavailability or failure to operate of any of the passive safety systems will change the accident sequence and may affect reactor safety. The analysis in this study is based on some hypothetical scenarios, in which the passive safety system failure is considered during the loss of coolant accidents. Four different cases are assumed, that is, with all passive systems, without actuation of one of the accumulators, without actuation of ADS stages 1–3, and without actuation of ADS stage 4. The actuation of all safety systems at their actuation set-points provides adequate core cooling by injecting sufficient water inventory into reactor core. The LOCA with actuation of one of the accumulators cause early actuation of ADS and IRWST. In case of LOCA without ADS stages 1–3, the primary system depressurization is relatively slow and mixture level above core active region drops much earlier than IRWST actuation. The accident without ADS stage 4 actuation results in slow depressurization and mixture level above core active region drops earlier than IRWST injection. Moreover, the comparison of cladding surface temperature is performed in all cases considered in this work. Anwar Hussain and Amjad Nawaz Copyright © 2016 Anwar Hussain and Amjad Nawaz. All rights reserved. Radiation Dose Calculations for a Hypothetical Accident in Xianning Nuclear Power Plant Tue, 01 Nov 2016 12:16:30 +0000 http://www.hindawi.com/journals/stni/2016/3105878/ Atmospheric dispersion modeling and radiation dose calculations have been performed for a hypothetical AP1000 SGTR accident by HotSpot code 3.03. TEDE, the respiratory time-integrated air concentration, and the ground deposition are calculated for various atmospheric stability classes, Pasquill stability categories A–F with site-specific averaged meteorological conditions. The results indicate that the maximum plume centerline ground deposition value of  kBq/m2 occurred at about 1.4 km and the maximum TEDE value of  Sv occurred at 1.4 km from the reactor. It is still far below the annual regulatory limits of 1 mSv for the public as set in IAEA Safety Report Series number 115. The released radionuclides might be transported to long distances but will not have any harmful effect on the public. Bo Cao, Junxiao Zheng, and Yixue Chen Copyright © 2016 Bo Cao et al. All rights reserved. Emergency Planning Zones Estimation for Karachi-2 and Karachi-3 Nuclear Power Plants using Gaussian Puff Model Sun, 30 Oct 2016 12:00:22 +0000 http://www.hindawi.com/journals/stni/2016/8549498/ Emergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K-2/K-3 NPPs) have been realistically determined by employing Gaussian puff model and Gaussian plume model together for atmospheric transport, diffusion, and deposition of radioactive material using onsite and regional data related to meteorology, topography, and land-use along with latest IAEA Post-Fukushima Guidelines. The analysis work has been carried out using U.S.NRC computer code RASCAL 4.2. The assumed environmental radioactive releases provide the sound theoretical and practical bases for the estimation of emergency planning zones covering most expected scenario of severe accident and most recent multiunit Fukushima Accident. Sheltering could be used as protective action for longer period of about 04 days. The area about 3 km of K-2/K-3 NPPs site should be evacuated and an iodine thyroid blocking agent should be taken before release up to about 14 km to prevent severe deterministic effects. Stochastic effects may be avoided or minimized by evacuating the area within about 8 km of the K-2/K-3 NPPs site. Protective actions may become more effective and cost beneficial by using current methodology as Gaussian puff model realistically represents atmospheric transport, dispersion, and disposition processes in contrast to straight-line Gaussian plume model explicitly in study area. The estimated PAZ and UPZ were found 3 km and 8 km, respectively, around K-2/K-3 NPPs which are in well agreement with IAEA Post-Fukushima Study. Therefore, current study results could be used in the establishment of emergency planning zones around K-2/K-3 NPPs. Sümer Şahin and Muhammad Ali Copyright © 2016 Sümer Şahin and Muhammad Ali. All rights reserved. Effects of a Mixed Zone on TGO Displacement Instabilities of Thermal Barrier Coatings at High Temperature in Gas-Cooled Fast Reactors Mon, 24 Oct 2016 06:52:51 +0000 http://www.hindawi.com/journals/stni/2016/9071237/ Thermally grown oxide (TGO), commonly pure α-Al2O3, formed on protective coatings acts as an insulation barrier shielding cooled reactors from high temperatures in nuclear energy systems. Mixed zone (MZ) oxide often grows at the interface between the alumina layer and top coat in thermal barrier coatings (TBCs) at high temperature dwell times accompanied by the formation of alumina. The newly formed MZ destroys interface integrity and significantly affects the displacement instabilities of TGO. In this work, a finite element model based on material property changes was constructed to investigate the effects of MZ on the displacement instabilities of TGO. MZ formation was simulated by gradually changing the metal material properties into MZ upon thermal cycling. Quantitative data show that MZ formation induces an enormous stress in TGO, resulting in a sharp change of displacement compared to the alumina layer. The displacement instability increases with an increase in the MZ growth rate, growth strain, and thickness. Thus, the formation of a MZ accelerates the failure of TBCs, which is in agreement with previous experimental observations. These results provide data for the understanding of TBC failure mechanisms associated with MZ formation and of how to prolong TBC working life. Jian Wang, Jun Ding, Kun Song, Song Chen, and Xia Huang Copyright © 2016 Jian Wang et al. All rights reserved. Development and Validation of a Three-Dimensional Diffusion Code Based on a High Order Nodal Expansion Method for Hexagonal- Geometry Mon, 17 Oct 2016 09:13:41 +0000 http://www.hindawi.com/journals/stni/2016/6340652/ A three-dimensional, multigroup, diffusion code based on a high order nodal expansion method for hexagonal- geometry (HNHEX) was developed to perform the neutronic analysis of hexagonal- geometry. In this method, one-dimensional radial and axial spatially flux of each node and energy group are defined as quadratic polynomial expansion and four-order polynomial expansion, respectively. The approximations for one-dimensional radial and axial spatially flux both have second-order accuracy. Moment weighting is used to obtain high order expansion coefficients of the polynomials of one-dimensional radial and axial spatially flux. The partially integrated radial and axial leakages are both approximated by the quadratic polynomial. The coarse-mesh rebalance method with the asymptotic source extrapolation is applied to accelerate the calculation. This code is used for calculation of effective multiplication factor, neutron flux distribution, and power distribution. The numerical calculation in this paper for three-dimensional SNR and VVER 440 benchmark problems demonstrates the accuracy of the code. In addition, the results show that the accuracy of the code is improved by applying quadratic approximation for partially integrated axial leakage and four-order approximation for one-dimensional axial spatially flux in comparison to flat approximation for partially integrated axial leakage and quadratic approximation for one-dimensional axial spatially flux. Daogang Lu and Chao Guo Copyright © 2016 Daogang Lu and Chao Guo. All rights reserved. Nonuniform Oxidation on the Surface of Fuel Element in HTR Tue, 11 Oct 2016 06:43:42 +0000 http://www.hindawi.com/journals/stni/2016/7485602/ The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR). The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600. Peng Liu, Yanhua Zheng, Wei Xu, and Lei Shi Copyright © 2016 Peng Liu et al. All rights reserved. Nondestructive Evaluation of Functionally Graded Subsurface Damage on Cylinders in Nuclear Installations Based on Circumferential SH Waves Thu, 29 Sep 2016 13:01:14 +0000 http://www.hindawi.com/journals/stni/2016/3035180/ Subsurface damage could affect the service life of structures. In nuclear engineering, nondestructive evaluation and detection of the evaluation of the subsurface damage region are of great importance to ensure the safety of nuclear installations. In this paper, we propose the use of circumferential horizontal shear (SH) waves to detect mechanical properties of subsurface regions of damage on cylindrical structures. The regions of surface damage are considered to be functionally graded material (FGM) and the cylinder is considered to be a layered structure. The Bessel functions and the power series technique are employed to solve the governing equations. By analyzing the SH waves in the 12Cr-ODS ferritic steel cylinder, which is frequently applied in the nuclear installations, we discuss the relationship between the phase velocities of SH waves in the cylinder with subsurface layers of damage and the mechanical properties of the subsurface damaged regions. The results show that the subsurface damage could lead to decrease of the SH waves’ phase velocity. The gradient parameters, which represent the degree of subsurface damage, can be evaluated by the variation of the SH waves’ phase velocity. Research results of this study can provide theoretical guidance in nondestructive evaluation for use in the analysis of the reliability and durability of nuclear installations. Zhen Qu, Xiaoqin Shen, and Xiaoshan Cao Copyright © 2016 Zhen Qu et al. All rights reserved. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor Tue, 27 Sep 2016 13:24:00 +0000 http://www.hindawi.com/journals/stni/2016/4385925/ The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor. Ki-Hwan Kim, Jong-Hwan Kim, Seok-Jin Oh, Jung-Won Lee, Ho-Jin Lee, and Chan-Bock Lee Copyright © 2016 Ki-Hwan Kim et al. All rights reserved. Numerical Investigation on Bubble Growth and Sliding Process of Subcooled Flow Boiling in Narrow Rectangular Channel Tue, 27 Sep 2016 09:54:04 +0000 http://www.hindawi.com/journals/stni/2016/7253907/ In order to investigate single bubble evolution, a boiling phase change model in subcooled flow boiling is proposed in this paper, and VOF model combined with phase change model is adopted to simulate the single bubble growth and movement. The effects of flow velocity, liquid subcooling, wall superheat, and vapor-liquid contact angle are considered in this model. The predicted bubble growth curve agrees well with the experimental result. Based on the analysis of bubble shape evolution and temperature field, it is found that the average bubble growth rate, flow velocity, and dynamic contact angle have significant effect on the bubble shape evolution during the bubble growth and movement while the temperature gradient in superheated liquid does not change with bubble growing. The character of dynamic contact angle during bubble growth and movement is also obtained in different working condition. De-wen Yuan, Zejun Xiao, Deqi Chen, Yunke Zhong, Xiao Yan, Jianjun Xu, and Yanping Huang Copyright © 2016 De-wen Yuan et al. All rights reserved.