bluuebunny/crossref_metadata_2025 huggingface dataset:
Heat Transfer and Fluid Flow Characteristic of U-Shaped Flow Channel for Applications of VHTR
Authors: Takayama Yasuaki , Takeda Tetsuaki
Published: October 2021

A very high-temperature reactor (VHTR) is a next-generation nuclear reactor systems. A gas cooling system with natural circulation is considered as a candidate for the pressure vessel cooling system (VCS) of the VHTR. The Japan Atomic Energy Agency is pursuing the design and development of commercial systems such as the 300 MWe gas turbine high-temperature reactor 300 for cogeneration (GTHTR300C). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV). A cooling panel utilizing the natural convection of air has been proposed. However, the amount of removed heat is inferior to that of cooling by forced convection. In this study, we use an experimental apparatus to simulate the cooling panel of a VCS. The experimental apparatus is a U-shaped flow channel, and the heating surface side is a vertical rectangular flow channel. Air is used as the working fluid. A fine copper wire is used as the porous material. The porosity is varied from 0.996 to 0.999. We perform an experiment to investigate the heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Additionally, we perform an experiment on a smooth surface for comparison.
Probabilistic Safety Assessment on Unavailability of Auxiliary External Power Supply in Fangjiashan Nuclear Power Plant
Authors: Zou Shengjia , Zhao Ming , Liu Deyi , Zhang Naiyuan , Luo Yang , Wang Li , Zhang Jianguo , Chen Honghao
Published: October 2021

The auxiliary external power Qinfang 2P70 of Fangjiashan NPP need to be shutdown, which makes one way the external power unavailable, thus affecting the safety level of the plant. In this paper, Through the internal event level PSA model, internal event LERF model, internal fire model and internal water flooded model, during Fangjiashan at power, 2P70 shut down caused one way the external power unavailable, gives the safety assessment results, And put forward the close attention of the important equipment such as the main power grid and the diesel engine. The test and maintenance of the diesel engine and the SEC/RRI heat exchanger are prohibited. The risk of the unit is managed and controlled by a series of other measures. The risk monitor is used to supervise the unit in real time. Finally, the application risk monitor of the power plant is discussed. The necessity and the PSA method to evaluate the possibility of auxiliary transformer shutdown to generalize to generating units.
SWIFT-RIMPUFF Modeling of Air Dispersion at a Nuclear Powerplant Site With Heterogeneous Upwind Topography
Authors: Dong Xinwen , Fang Sheng , Zhuang Shuhan
Published: October 2021

The SWIFT-RIMPUFF can provide refined atmospheric dispersion modeling for nuclear emergency response, but its performance for the mesoscale range in a nuclear power plant (NPP) site with highly complex topographies hasn’t been fully investigated. In this study, a validation of SWIFT-RIMPUFF was performed based on a wind tunnel experiment simulating a real China’s multi-reactor NPP site with heterogeneous upwind topography and dense buildings to understand the potential discrepancies or limits. The results demonstrate that the SWIFT-RIMPUFF can reproduce the sharp changes of wind flows for both speed and directions near the buildings, but usually overestimate the wind speed in the complex topography. For vertical wind profiles, the accuracies show high dependencies on the local topography and building layout, and the deviation of those near the building is more obvious. The simulated ground concentrations match the topographic changes of high-altitude mountains. The concentration predictions in the downwind building area are acceptable which displays that the influence of building effects can be well introduced, but the simulations in the building area still show noticeable discrepancies when compared with those in the sea area. However, such deviations do not propagate to the downwind mountainous and sea areas, which the accuracies are quite satisfactory.
Spreading Behavior of Molten Metal on Flat Plate in a Shallow Water Pool
Authors: Yamamoto Yasunori , Ito Tomomasa , Nihashi Kyosuke , Miwa Shuichiro
Published: October 2021

The spreading behavior of the molten core is an essential phenomenon to consider for understanding the initial conditions of molten core-concreate interaction and the coolability of the core. In this study, experiments using molten tin were performed to obtain the necessary data to analyze the spreading behavior of the molten core under wet conditions. The experimental parameters used were the nozzle diameter, water depth, and molten metal temperature. The pressure waves occurred due to the molten metal-water interaction in the water pool, especially under high molten metal temperature conditions. They were observed during the floor collision process and the spreading process after the collision. Our results indicated that the spreading area and the shape of the solidified metals strongly depended on the pressure waves due to the thermal interaction between the molten tin and water. When the interface temperature between the molten tin and water was lower than the homogeneous nucleation temperature, the pressure wave rarely occurred. The effect of the pressure wave on the spreading areas increased with the superheat of the molten tin.
Theoretical Study on Single Bubble Expansion After Pressurized Subcooled Water Injected in Thermal Glycerin
Authors: Mao Feng , Zhang Lei , Yun Xiangyu , Lu Donghua , Tian Wenxi , Zhang Huiyong
Published: October 2021

In the hypothetical liquid metal reactor SGTR (Steam Generator Tube Rupture) accident, the high-pressure water in steam generator tube will be injected into the high-temperature liquid metal. During the water-liquid metal interaction, water vaporized and vapor bubbles migrated into liquid metal, which could be resulting a sharp pressure raise and threatening the safety of the reactor. This multi-phase system can be abstracted as a system in which a high-pressure subcooled fluid (overheating state at atmospheric pressure) is injected into another higher temperature fluid (normal pressure) to form a two-component (water and liquid metal) system in which multiphase (fluid and vapor) were existed.

There are two effect could cause bubbles expansion which lead to overall system pressure increase, they are flash evaporation of pressurized water since the depressurization and high-temperature glycerin heating the water/vapor. In this paper, the theoretical analysis of the single bubble expansion model is used to evaluate the effect of these two phenomena on the system pressure rise. First of all, this paper uses the force balance point of view to derive the single bubble expansion equation inside the inviscid fluid, and obtain the relationship between bubble expansion speed and bubble internal pressure. Secondly, through conservative assumptions, the bubble expansion speed caused by the heating effect of the high temperature fluid on the bubble is evaluated. Finally, this paper compares the strength of these two effects in the expansion of a single bubble, and theoretically explains the driving force of the pressure increase in the simulation experiment and the potential flow pattern.
Study on Aging Management of Operating Nuclear Power Plants in China
Authors: Liang Li , Guixiang Yi
Published: October 2021

In 2016, The first nuclear power plant submitted an application for renewal of operation license to National Nuclear Safety Administration (NNSA) in China. In order to guide the project, NNSA issued the technical policy for extension of the validity period of the nuclear power plant operation license, which mainly refers to the US nuclear power plant license renewal review policy. However, China’s nuclear safety regulation system mainly refers to the establishment of the International Atomic Energy Agency (IAEA), and carries out 10-year Periodic Safety Review for nuclear power plants in the actual operation process. There are some differences between the U.S. license renewal strategy and the IAEA long-term operation system. An important part of both is aging management. Based on the daily operation and supervision of China’s nuclear power plants, this paper summarizes the differences and similarities of aging management between periodic safety review and operation license renewal application, and puts forward the contents that should be added and improved in the periodic safety review, so as to effectively link up the application system for operation license renewal and the Periodic Safety Review system, to revise the “Periodic safety review of nuclear power plant” (HAD103/11) and “Aging management of nuclear power plant” (HAD 103/12) provide reference.
Preliminary Design of a Fuel Element With Divergent Hot Gas Channel In Particle Bed Reactor for Nuclear Thermal Propulsion
Authors: Liang Zhaoyu , Ji Yu , Sun Jun , Mao Chenrui , Shi Lei
Published: October 2021

The nuclear thermal propulsion (NTP) system can shorten the travel time in deep space exploration and reduce the initial mass of the launch vehicle due to its superior characteristics including high specific impulse and large thrust. Particle bed reactor (PBR) is one of the most appropriate reactor concepts to equip the NTP systems. To make the best use of PBR, the thermal-hydraulic design of the fuel element should be carefully considered and a flow-power matching technology should be developed. In this paper, a novel design employing a divergent hot gas channel is proposed to achieve a uniform flow distribution with lower maximum temperature and pressure drop. Through the analysis of the 1D modified momentum equation in the inlet plenum and hot gas channel, the model of pressure drop is established. Then, the differential equation of the ideal cross-section of the hot gas channel is derived. At last, the flow and heat transfer process in the fuel element with divergent hot gas channel is simulated by using computational fluid dynamics (CFD) code, and the reduction of pressure drop and temperature verifies the theoretical model.

This study shows that the proposed design of the divergent hot gas channel can provide a new idea for thermal-hydraulic optimization of the PBR fuel element.
Preliminary Research on The Thermal-Mechanical Coupling Behavior Simulation Method of M3 Fuel
Authors: Changbing Tang , Yongjun Jiao , Yuanming Li , Kun Zhang
Published: October 2021

M3 fuel which microencapsulated TRISO particles in zirconium metal matrix is an integrated fuel designed to replace the traditional pellet-cladding fuel rod. The fuel design can effectively solve the problems of LWR fuel, such as water side corrosion, abrasion, pellet-cladding interaction, stress corrosion cracking and internal pressure bracing. Because of the multiple coating structure of TRISO particles and complex contacting behavior and deformation between TRISO particle and metal matrix, M3 fuel exhibits complex in-pile behavior. In order to establish the simulation method for the M3 fuel, the three-dimensional model was established by using ABAQUS software. The complex contacting behavior and deformation of coated layers and matrix for M3 fuel was considered. The results indicated that bonded strength between TRISO particle and Zr matrix has great effect on the stress distribution of SiC layer and matrix. The vulnerable area of Zr matrix located in the nearest position of TRISO particle to TRISO particle. The maximum tensile hoop stress of SiC layer was up to 150 MPa under strong bonding state, while SiC layer suffered compressive stress during all operation time for the weak bonding state. The failure probability of SiC layer was about 7 × 10−3 under strong bonding state which much higher than the weak ones. Thus, the integrity of SiC layer can be maintained by controlling the bonding strength between TRISO particle and Zr matrix during manufacturing process.
A New Insight Into Molten Corium Concrete Interaction With Concrete Ablation Analysis for Mitigation Scheme
Authors: Khurshid Ilyas , Alade Amidu Muritala , Addad Yacine , Afgan Imran
Published: October 2021

The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, composition of concrete, molten corium heat transfer models and solve the related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH were successfully utilized which enabled the modeling of interaction between molten corium and concrete. The developed model was validated against experimental data and the results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating depth of ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete is the main controlling factor to mitigate ablation in the investigated case study. An alternative to concrete, could be a certain igneous rock (tested in this study), can lead to comparatively low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. As a consequence of concrete ablation by molten corium, the depth of ablation is expected to decrease to zero ablation with the suggested material in the selected case study. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperature, respectively.
Exploring Probability of Gas Entrainment With CFD Analysis of the Flow in the MICAS Experimental Facility
Authors: Bhatia Harshit , Bieder Ulrich , Gorsse Yannick , Guenadou David
Published: October 2021

Investigation of sodium cooled fast reactors (SFR) is ongoing in France. There have been a number of experiments carried out within the framework of Generation IV reactor program at CEA. MICAS experimental facility is one of these experiments, which represents the hot pool facility of a SFR in a 1/6th scaled down model. Recent studies carried out in MICAS, explore the probability of vortex formation at one of the internal heat exchangers (IHX) nozzle, which can further lead to gasentrainment from the pool surface. Therefore, MICAS facility has extensively tested different flow conditions and the non-symmetric IHX operation.

Multi-scale calculations are carried out using Computational Fluid Dynamics (CFD) in order to better understand the vortex formation at the IHX plenum and the associated possible gas entrainment into the vortex from the hot pool free surface. This paper presents the computational analysis with focus on operation of the SFR, which tries to replicate the possibility of vortex formation, under non-symmetric operation of two out of four IHXs. The analysis was carried out with the open source code TrioCFD. This study provides valuable insight into possibilities, constrains and limitation of the used CFD calculation methodology.
The Evaluation of Break Sizes of LOCA by the Temperature Difference at the Recirculation Inlets of BWR
Authors: Hwang Sheng-Dih
Published: October 2021

In this study, the MAAP5 code was used as a tool to simulate a BWR6/MarkIII nuclear reactor and which was LOCA with different break sizes at the recirculation loops. By detecting the temperature difference between the two loops, and then the break size could be determined. The results were consistent with the previous experiments or simulations. The study could be suggested to the operators at the main control room (MCR), when they observed some syndromes of the data and they could preliminary estimate the size of the LOCA. Those could help the operators to determine a strategy to avoid the probability of severe accident.
Development of an Integrated Design Evaluation ‘HITEP’ Platform for High-Temperature Pressure Boundary Components and Piping Systems
Authors: Lee Hyeong-Yeon , Jeong Si-Hwa , Won Min-Gu , Huh Nam-Su
Published: October 2021

An integrated software platform of elevated temperature design evaluation for pressure boundary components and piping subjected to high-temperature operation in the creep range has been developed. The software of ‘HITEP’ platform is capable of performing high-reliability design evaluations according to two nuclear grade design rules of ASME BPVC Section III Subsection HB and RCC-MRx, and one non-nuclear grade the Section VIII Division 2 Code Case 2843-2. The program modules were verified with a number of example problems. The HITTEP platform can be used for a reliable and efficient elevated temperature design of pressure boundary components and piping systems in generation IV nuclear systems, supercritical thermal power plants and oil refinery plants subjected to high-temperature operation in the creep range.
Heat Transfer Analysis of a Conceptual Horizontally-Oriented High Temperature Gas-Cooled Reactor
Authors: Feng Jinyong , Lopez Enrique Velez , Stewart William Robb , Wiser Ralph , Baglietto Emilio , Shirvan Koroush
Published: October 2021

High temperature gas-cooled reactors (HTGRs) have been constructed around the world since the 1960s. Compared to light water reactors, HTGRs feature a low core power density, which ultimately results in the need for larger components and structures. The successful widespread deployment of HTGRs will likely require a significant reduction in reactor building size per kW, to compete with other sources of carbon-free energy. The Modular Integrated Gas-cooled High Temperature Reactor (MIGHT-R) has been recently proposed to leverage inherent safety characteristics of HTGRs and its proven technology, while increasing its reactor building power density by 4 to 5 times. Contrarily to the vertical side-by-side orientation of the reactor core and steam generator of the typical HTGR design, the MIGHTR primary system components are laid horizontally, inline with one another. In this paper, a design embodiment of MIGHT-R concept is presented and preliminary heat transfer analysis under normal operating conditions is performed adopting a simplified porous representation of the flow behavior inside the reactor core and steam generator. The predicted mass flow rates in the fuel region, inner reflectors and outer reflectors agree with analytical estimations. A set of sensitivity tests for components variations are presented to provide insights on the influence of design parameters. A preliminary design reactor cavity cooling system (RCCS) is presented which is basically a 100 °C “water shell” surrounding the reactor vessel and it only covers the reactor core regions. Comparison of the same design with different orientations is also made and indicates that the mass flow and temperature distribution inside the reactor core have low dependence on orientation. Based on the computational results, the stratification of the coolant in the horizontal configuration for normal operation is deemed negligible. The presented work supports the normal operation feasibility of a horizontally-oriented HTGR concept with its potential to significantly reduce HTGR capital cost.
Failure Strain Criterion Accounting for the Effect of Material Strength
Authors: Baghous N. , Barsoum I.
Published: October 2021

The objective of this study is to investigate the effect of the Lode parameter on different material strengths. Recent work has shown that ductile failure highly depends on the stress state characterized by both the stress triaxiality T and the Lode parameter L, which is related to the third deviatoric stress invariant. Thus, for six different steel grades, two different specimen geometries were manufactured to account for two different Lode parameters (L = −1 and L = 0), whereas T is controlled by introducing different sized notches at the center of the specimens. By performing tensile experiments and running finite element simulations, the ductile failure loci of the six materials showed variations between the two specimen geometries, indicating that the failure highly depends on the stress state characterized by both T and L. This indicates the need to reassess the ductile local failure criterion in the ASME codes that only accounts for T as a stress state measure. A Lode sensitivity parameter LS is defined based on the experimental results and revealed that the steel grades with ultimate strength higher than a certain threshold value (450 MPa) exhibit sensitivity to the Lode parameter, and the results showed that the LS increases with increase in the ultimate strength of the steel grade. The results were incorporated to enhance the original ASME local failure criterion by accounting for T, L, and LS to accurately assess ductile failure in high-strength steels.
Application of Ultrasonic Guided Wave in LMPH Tube of Ethylene Cracking Furnace
Authors: Ding Ju , Wang Shenghui , Liu Shuhong
Published: October 2021

In view of the frequent leakage and perforation of the steam system pipe fittings of a Chinese factory in a short period of use, it will not only affect the normal use of the pipe system, but also cause safety hazards. In this paper, sampling and analysis of failed pipe fittings are carried out. Through the analysis of macro and micro corrosion morphology on site, it is shown that local corrosion is the root cause of failure, and it is accompanied by uniform corrosion. When the corrosion reaches a certain level, such as perforation, the overall pipeline will increase larger system security risks. So, what can we do to prevent these accidents? In this article, we make some useful suggestions: Strict control of the quality of cooling water. Do a good job of steam defogging. Timely replace the pipe fittings with serious thinning and corrosion thinning. Change the material to ultra-low carbon stainless steel (316L) or Wear-resistant materials. Comprehensive inspection and structural integrity safety assessment to ensure the safe and effective operation of equipment.
Technical Basis for Proposed ASME Code Case for Construction of Compact Heat Exchangers in High Temperature Reactors
Authors: Keating Robert , McKillop Suzanne
Published: October 2021

The U.S. Department of Energy has recently completed a research program to support the development of the compact heat exchanger (CHX) for use in high temperature advanced reactors. The project was executed by an Integrated Research Project (IRP) and includes team members from the University of Wisconsin–Madison, University of Michigan, Georgia Institute of Technology, University of Idaho, North Carolina State University, Oregon State University, Electric Power Research Institute, MPR Associates, and heat exchanger manufacturers CompRex and Vacuum Process Engineering.

The research was guided by a roadmap to support the eventual adoption of an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Code Case for compact heat exchanger designs in high temperature reactor service. This roadmap is outlined in Journal of Nuclear Engineering and Radiation Science, NERS 19-1161, which identified specific research gaps to support the Code Case. The IRP has completed the research and the ASME Section III Standards Committee is in the process of writing and approving a Code Case for Section III, Division 5, Class A applications.

This paper provides the detailed technical basis for the ASME Code Case for CHX. The paper provides a vital link between the research results and the proposed Code Case and closes the research gaps that are tied to the original Roadmap. In addition, the paper provides further detailed commentary on the proposed requirements. The major technology gaps included material properties, failure modes and effects, analysis methods, and examination methods. The IRP studied basic material properties of diffusion bonded plate, creep and fatigue models, development of NDE methodology and development of advanced analytical approaches to design.
Prediction Method and Application Case Analysis of Dangerous Parts of Pipelines in Petrochemical Plants
Authors: Fang Zhou , Chen Ruxin , Wang Zhe , Huan Zhibo , Liu Qia , Yan He
Published: October 2021

A prediction method of dangerous points resulting from dynamic erosion was studied by combining Computational Fluid Dynamics and application cases. CFD numerical simulation software was used to simulate conditions of real process in industry. Considering the factors like geometry, flow velocity, flow direction, different erosion results in various situation were exampled and analyzed. Working conditions were sorted out and related sample graphs were made and indexed. This prediction method can reveal flow law of internal flow field and flow field distribution. Hence, critical points caused by dynamic erosion can be found and sensor-layout plan for online inspection, monitoring and management can be introduced. The safety in the use of industrial pipelines can be guaranteed. This method had been applied on pipelines of an oil and gas company in western China. A series of studies regarding their erosion rates and trend predictions were done based on actual situation on site. The technical route for applying this method was explained in detail. The corrosion mechanism and calculation method for this case were concretely stated. The prediction results and influencing factors of erosion parts were stated and the significance of using above-mentioned prediction method is declared.
Application of Different Data-Driven Methods in Material Performance Prediction of 2.25Cr-1Mo-0.25V Steel After Forming and Tempering
Authors: Li You , Chen Zhiping , Li Hongfei
Published: October 2021

2.25Cr-1Mo-0.25V steel is the main material of hydrogenation reactor due to its good high temperature properties and hydrogen embrittlement resistance. As we all know, there is a great performance gap between virgin material and actual product. Nevertheless, current design methods cannot evaluate the impact of manufacturing. In engineering practice, tensile testing is widely implemented on online surveillance samples or experimentally simulated samples to acquire the actual material characteristics. However, sampling from service hydrogenation reactor, as well as simulating actual manufacturing process, is time-consuming, labor-intensive, and difficult. To systematically analyze the effect of fabrication on 2.25Cr-1Mo-0.25V steel, a prediction method that can consider manufacturing residual effects needs to be proposed. Nowadays, many data-driven methods, such as back propagation neural network (BPNN), support vector machine (SVM) and random forest (RF), are employed in predicting material behavior during hot forming, whereas they are scarcely applied on 2.25Cr-1Mo-0.25V steel. In this work, the mentioned three machine learning procedures were constructed to simulate the behavior of actually manufactured 2.25Cr-1Mo-0.25V steel in uniaxial tensile test. Firstly, tensile samples were extracted from different locations of an actually manufactured hydrogenation reactor, and stretched at 454 °C and 482 °C. Subsequently, BPNN, SVM and RF models were developed on the obtained tensile curves, respectively. Their inputs were manufacturing parameter, sampling location, service temperature and tensile strain, whereas their output was tensile stress. Finally, the predictability of models was evaluated by test set data which were not involved in model training process. The results show that the predictions of BPNN, SVM and RF all agree well with experimental data. BPNN exhibits lowest prediction error and highest generalization power, SVM is somewhere in between, while RF is the worst. This work will contribute to the application of data-driven methods in material property prediction of 2.25Cr-1Mo-0.25V steel.
Experimental Investigation on the Seismic Performance of a Multi-Component System for Major-Hazard Industrial Facilities
Authors: Butenweg Christoph , Bursi Oreste S. , Nardin Chiara , Lanese Igor , Pavese Alberto , Marinković Marko , Paolacci Fabrizio , Quinci Gianluca
Published: October 2021

Past earthquakes demonstrated the high vulnerability of industrial facilities equipped with complex process technologies leading to serious damage of the process equipment and multiple and simultaneous release of hazardous substances in industrial facilities. Nevertheless, the design of industrial plants is inadequately described in recent codes and guidelines, as they do not consider the dynamic interaction between the structure and the installations and thus the effect of seismic response of the installations on the response of the structure and vice versa. The current code-based approach for the seismic design of industrial facilities is considered not enough for ensure proper safety conditions against exceptional event entailing loss of content and related consequences. Accordingly, SPIF project (Seismic Performance of Multi-Component Systems in Special Risk Industrial Facilities) was proposed within the framework of the European H2020 - SERA funding scheme (Seismology and Earthquake Engineering Research Infrastructure Alliance for Europe). The objective of the SPIF project is the investigation of the seismic behaviour of a representative industrial structure equipped with complex process technology by means of shaking table tests. The test structure is a three-story moment resisting steel frame with vertical and horizontal vessels and cabinets, arranged on the three levels and connected by pipes. The dynamic behaviour of the test structure and of its relative several installations is investigated. Furthermore, both process components and primary structure interactions are considered and analyzed. Several PGA-scaled artificial ground motions are applied to study the seismic response at different levels. After each test, dynamic identification measurements are carried out to characterize the system condition. The contribution presents the experimental setup of the investigated structure and installations, selected measurement data and describes the obtained damage. Furthermore, important findings for the definition of performance limits, the effectiveness of floor response spectra in industrial facilities will be presented and discussed.
Study on Reliability of Seismic Capacity Analysis for Important Equipment in Nuclear Facilities
Authors: Muta Hitoshi , Ohtori Yasuki , Furuya Osamu , Fujita Satoshi
Published: October 2021

Since the severe accident of Fukushima Dai-ichi Nuclear Power Plants, the enhancement and improvement of the safety regulations of nuclear facilities in Japan from various perspectives have been considered the new regulations are based on defense in depth concept as well as before Fukushima Accident, and have significantly raised design basis assumptions and required countermeasures for natural phenomena to prevent the simultaneous loss of safety functions due to the dependency.

From the above background, according to raised design basis assumptions, Severe Accident (SA) countermeasures against beyond design basis external hazards including earthquakes, tsunami and so on have been considered and the plans of installation of them have been accepted by Nuclear Regulatory Authority. On the other hand, since Probabilistic Risk Assessment (PRA) and Safety Margin Analysis (SMA) are performed in the safety improvement evaluation to ensure the residual risk of severe accident and confirm the actual performance of safety functions. However, since less information could be obtained from the experienced capacity data or operation experiences, the safety insights from PRA and SMA have not been clarified so far.

This paper presents the research and examination results of the reliability of capacity data based on the very few experience data by using Statistical Technique and Bayesian Update Technique to enhance the reliability and validity of PRA and SMA and summarizes kinds of insights obtained by this research and examination.
A Simplified Method for Analyzing Pressure Vessel Response to An External Blast Overpressure Generated From an Explosion
Authors: Hundl Jacob , Millet Barry , Mosher Bryan , Kirkpatrick Kenneth
Published: October 2021

The widespread increase in the construction of liquefied natural gas (LNG) processing facilities and gas producing offshore platforms has produced new requirements in designing equipment, in this case pressure vessels, to withstand a blast overpressure condition resulting from the failure of an LNG train or some other form of gas-based explosion. New client requirements are aimed at determining the viability of equipment after one of these catastrophic events. Traditional methods of analyzing pressure vessel response to blast overpressure, such as finite element analysis (FEA) and computational fluid dynamic (CFD) studies, are costly and impractical when applied to large scale projects and tens or hundreds of pieces of equipment. To adequately account for and measure pressure vessel response to blast overpressure, a simplified analytical method is required. However, a simplified method that does not consider the unique conditions presented by an overpressure event produce extremely conservative results (e.g. applying the overpressure as a wind load or pressure times area). Unfortunately, there is limited methodology and literature on blast overpressure phenomena, particularly related to petrochemical refining processes.

This paper will examine the development of a simplified method for analyzing pressure vessel response to blast overpressure using methods originally intended for the study of the effects of nuclear weapons on structures as well as more traditional methodology which employs an explicit solution utilizing modern, commercially available FEA software. The paper will review the methods presented in the existing literature and examine how those methods can be efficiently and practically applied to pressure vessels. The results generated by the simplified methodology will be compared to results from a control case that uses an explicit FEA method to evaluate pressure vessel response to blast overpressure.
Statistically Analyzing the Characteristics of Gas Pipelines Accidents Based on PHMSA Database
Authors: Liu Shengli
Published: October 2021

Deeply analyzing the temporal characteristics of gas pipeline accidents is crucial to the safety management of gas pipeline. In this study, gas distribution pipeline accidents recorded from 1986 to 2020 provided by PHMSA have been chosen to explore the characteristics of gas pipeline failure frequency in the United States. The result indicates that the failure frequency of gas pipelines has been reduced with the improvement of safety management. The analysis results show that the gas pipeline accidents occurred in the U.S. from 1986 to 2020 are temporal connected. This is due to the fact that the leading failure causes are related with human factors, which have strong temporal correlations. Improving the pipeline safety management can reduce the temporal dependence. However, there are essential temporal pattern differences among gas pipeline accident sequences of different severity. Accurate prediction of the gas pipeline accident occurrence can improve the quantitative risk assessment of gas pipeline accidents, and also help the pipeline management personnel to develop an efficient emergency response plan. Therefore, as for the quantitative risk assessment of gas pipeline accidents, it is necessary to carry out in-depth analysis of the occurrence probability according to the severity of the accidents.
A Numerical Approach to Correlate Compression Stress Relaxation and Compression Set of Elastomer O-Rings With Tightness
Authors: Weber Mike , Zaghdoudi Maha , Kömmling Anja , Jaunich Matthias , Wolff Dietmar
Published: October 2021

The excellent mechanical properties of elastomer seals at a wide range of temperatures as well as their high versatility and recovery potential under several load conditions make these materials well suitable for the application in containers designed for transport and disposal of negligible heat generating radioactive waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered which prohibit an avoidable cask handling. An extensive knowledge of the change of the elastomer properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the safe enclosure of the radioactive material for the required time are mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have already been made and are still ongoing to scientifically support this task.

Among other representative types of elastomers, specimen made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the with respect to application most important of their complex mechanical properties.

Exemplary results of these investigations were used to calibrate material models implemented in the commercial finite element software ABAQUS/Standard®. The finite element model already presented in previous works uses a sequential temperature displacement coupling. The calculated compression stress relaxation (CSR) and compression set (CS) values do satisfactorily match the experimental results.

In many investigations performed at BAM both values (CSR and CS) were identified as key indicators of elastomer’s long-term performance.

However, the possibility to correlate these equivalent indicators with performance values such as tightness and leakage rate, measurable in the mounted state, is an important goal of our future work. In the presented study the ABAQUS® feature of “pressure penetration” is introduced in the suggested finite element model for this purpose. It provides the possibility to simulate the penetration of a gas into a possible gap between flange and O-ring causing an opening of a leakage path. Three dimensional and axis-symmetric finite element models were generated to represent flat and grooved flanges of different dimensions.

The sensitivity of the feature to several input parameters is investigated and the observed behavior of the O-ring is correlated with the results of performed leakage tests.
Applicability of Design Rules for Openings in Shells, ASME B&PV Code, Section VIII, Division 2 - A Case Study
Authors: Kagita Gurumurthy , Pudipeddi Krishnakant V. , Sripada Subramanyam V. R.
Published: October 2021

The Pressure-Area method is recently introduced in the ASME Boiler and Pressure Vessel (B&PV) Code, Section VIII, Division 2 to reduce the excessive conservatism of the traditional area-replacement method. The Pressure-Area method is based on ensuring that the resistive internal force provided by the material is greater than or equal to the reactive load from the applied internal pressure. A comparative study is undertaken to study the applicability of design rules for certain nozzles in shells using finite element analysis (FEA). From the results of linear elastic FEA, it is found that in some cases the local stresses at the nozzle to shell junctions exceed the allowable stress limits even though the code requirements of Pressure-Area method are met. It is also found that there is reduction in local stresses when the requirement of nozzle to shell thickness ratio is maintained as per EN 13445 Part 3. The study also suggests that the reinforcement of nozzles satisfy the requirements of elastic-plastic stress analysis procedures even though it fails to satisfy the requirements of elastic stress analysis procedures. However, the reinforcement should be chosen judiciously to reduce the local stresses at the nozzle to shell junction and to satisfy other governing failure modes such as fatigue.
Comparative Research on Nonlinear Analysis Method Technique for the Failure Modes Based on RCC-M Annex ZC
Authors: Shao Xuejiao , Xie Hai , Xiong Furui , Fu Xiaolong , Tian Jun , Zhang Ying , Shi Kaikai , Yu Mingda
Published: October 2021

In the evaluation of nuclear equipment, if the elastic analysis cannot meet the corresponding design criteria, the nonlinear analysis should be adopted to assess the safety of equipment. A new nonlinear calculations appendix is supplied in RCC-M 2017 edition annex ZC to secure the nonlinear F.E. calculation performance.

In this paper, the failure modes of nuclear engineering structure are discussed. In order to ensure the safe operation of the structure, the corresponding evaluation criteria formulated by RCC-M code are proposed. Based on the annex ZC of RCC-M 2017 edition, three kinds of nonlinear analysis methods, including limit analysis, plastic analysis and cycle analysis, are studied. The austenitic stainless steel typical nozzle model is used to carry out the finite element calculation. The effect on the limit load is compared between the limit load method of perfect elastoplastic analysis and the secant method of the elastoplastic analysis. The influence of the strain range method and the modified Ke method on the fatigue calculation results is also discussed. The strain range can be determined by monotonic calculation or cyclic calculation. The two kinds of approach are implemented in the elastic plastic fatigue calculation. Compared with the simplified elastic plastic fatigue results, using Ke method to calculate elastic-plastic fatigue can reduce the conservatism of the results.
Fatigue Crack Growth Rate Data Assessment
Authors: Wang Enyang D. , Tiku Sanjay , Dinovitzer Aaron
Published: October 2021

A comprehensive fatigue test database was developed in PRCI SIA-1-1 and IM-3-2 projects, where there are 185 sets of fatigue crack growth rate parameters (i.e., C and m fitting coefficients in the Paris law) calculated from compact tension (CT) fatigue test data. Four sets of recommended values of C and m were presented for two flaw locations (base metal (BM) and Electric Resistance Weld seam (ERW)) and two R ratios (R < 0.5 and R ≥ 0.5). The past works did not further assess the fatigue crack growth rate parameter relation (i.e., C and m), although intuitively, two parameters should be correlated for a material type (i.e., line pipe). To this end, a systematic study was carried out to reassess the fatigue crack growth rate database. It is found that parameters C and m are strongly correlated in the form of m = A * log(C) + B, where A and B are fitting coefficients. In addition, C and m (mresidual) values follow lognormal and normal distributions, respectively, where mresidual refers to the residuals associated with fitting the predicted values. The values of A and B for BM or ERW at R = 0.1 or 0.6 (four scenarios) are reported. Note that the distributions of C and m for ERW cases have relatively heavier tails and shorter peaks.

Based upon the four newly developed sets of A and B values, a series of probability studies were carried to calculate the most-likely fatigue crack growth life in the integrity analysis. Note that two ERW pipe segments reported in PRCI IM-3-2 projects were considered in the full-scale test, where fifty-one (51) initial notches were prepared to evaluate the fatigue crack growth rate. The initial notches were cycled to develop different crack lengths, depths, materials, and flaw locations (e.g., BM and ERW seam). The fatigue crack growth lives calculated from the PRCI approach were used as the baseline for comparisons. In total, 30,000 pairs of C* and m* data points were generated from random sampling for each scenario, and there are over 1.5 million random simulations carried out. The probability densities for each full-scale test case were simulated using a Monte-Carlo approach. The fatigue crack growth life distributions for BM are relatively narrow because of the fatigue crack growth rate parameters’ lower standard deviation. The most-likely (median) fatigue life from the probabilistic study is almost identical to the fatigue life from deterministic calculations for the BM cracks. The probabilistic simulation results of ERW seam crackfatigue lives are significantly different from those for BM cracks.
Inservice Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part II: NDE/Sensor Technology Development and Codification
Authors: Meyer Ryan M. , Renshaw Jeremy , Hunter Kenn , Orihuela Mike , Stadler Jim , Mauskar Karan , Wolf Uwe , Klein Marvin , Weidmann Max , Bobbs Bradley , Dunn Darrell , Lin Bruce
Published: October 2021

This paper describes development and demonstration of nondestructive examination (NDE) technologies to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA to verify continued safe operation and that the canister confinement is intact and performing its intended safety function. Specifically, this work relates to NDE technology development for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond the initial license or certified term. Physical access to the canister surface is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics related to inspections. Several activities to demonstrate NDE technologies for the inspections of different DSS systems are summarized.
Case Study to Evaluate Stresses in Welded Endplates in Structural Steel
Authors: Gawande Kshitij P. , Mayes Alex , Subedi Raju
Published: October 2021

Endplates are widely used in the industry to attach supplementary steel structures to main building frames. These endplates can be attached to the building steel using a bolted connection or a welded connection. Industry often favors bolted connections due to ease of installation and availability of qualification methods per AISC 360 Design Guides. However, there are some applications where a welded connection is preferable, such as, cases requiring reduction of number of parts supplied or applications with higher chance of vibration causing loosening of bolts. The present case study discusses evaluation of stresses in welded endplates due to forces and moments from the attaching supplementary steel members. The study considers various welded connection scenarios including an endplate welded on two opposite sides and an endplate welded on all four sides. The stress distribution in the plate is studied using finite element analysis with wide flange and tube steel members attaching to it. ANSYS mechanical is used to perform the finite element analysis. Multiple combinations of plate sizes, weld patterns, and attaching member sizes are analyzed to provide a well-rounded solution. An analytical model is developed for the stress evaluation as well and the results are compared with the finite element model. The study is intended to provide an efficient methodology for plate evaluation and qualification.