bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Systematic Study of the Material Performance of Hot Isostatically Pressed Type 316L Stainless Steel Powder for the Civil Nuclear Sector
Authors: Kyffin William , Gandy David , Burdett Barry 
Published: October 2018

Hot Isostatic Pressing (HIP) of type 316L stainless steel powder has been an established manufacturing practice for more than twenty-five years in the oil and gas sector and more recently in the naval defence sector. To demonstrate the capability of the powder metallurgy HIP (PM/HIP) for nuclear power applications a systematic study of 316L commercial powder production, encapsulation/consolidation providers and selected HIP parameters was undertaken by the Nuclear AMRC in collaboration with the Electric Power Research Institute (EPRI). In the study, the 316L powder specification limited the oxygen content of the powder to under 130 parts per million (ppm), which reflects the improvements that commercial powder suppliers have been making over the past decade to ensure greater powder cleanliness.
The test programme assessed powder supply, HIP service provider and HIP sustain time. Excellent test results were achieved across the full range of variables studied with all billets meeting the specification requirements of ASTM A988 and additional requirements imposed based on nuclear manufacturing standards.
Significantly, the study demonstrated the robustness of the PM/HIP supply chain, as material produced via differing HIP service providers resulted in very consistent material properties across the destructive test programme. Furthermore, no significant difference in material properties were noted for material HIP’ed between 2–8 hours hold time, suggesting that the HIP process window is large. Both these results are significant from an end-user standpoint as they highlight the uniformity of the process through the full manufacturing cycle from powder procurement to destructive testing.
Despite all material passing specification requirements, some property variation was noted for differing powder suppliers. Considering the systematic approach, this was attributed to powder composition, with both low oxygen and high nitrogen contents contributing to improvements in Charpy impact strength and tensile strength respectively.
Analysis of Major Group Structures Used for Nuclear Reactor Simulations
Authors: Di Filippo Marco , Krepel Jiri , Mikityuk Konstantin , Prasser Horst-Michael 
Published: October 2018

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation.
The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated.
The aim was to identify a group structure with very low dependency on the actual reactor spectrum.
Experimental Study on the Steam Line Break (SLB) Accident With the Steam Generator Tube Rupture (SGTR)
Authors: Park Yusun , Bae Byoung Uhn , Kim Jongrok , Lee Jae Bong , Park Hae Min , Choi Nam Hyun , Kang Kyoung Ho , Kim Kyung Doo 
Published: October 2018

The integral effect test to simulate a steam line break (SLB) accident accompanied by a steam generator tube rupture (SGTR), named SLB-SGTR-02 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident.
In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomena which can occur in the nuclear power plant, especially for the pressurized water reactor, as realistically as possible. The SLB is simulated as a guillotine break on the main steam line and, in the SGTR, five steam generator tubes on the steam generator hot-side are modeled. Due to the initiation of SLB, a reactor was tripped by low steam generator pressure (LSGP) signal and the SGTR was initiated when the water inventory in the secondary system of the affected steam generator was dried-out.
During the transient simulation for SLB-SGTR-02 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops and the fluid temperatures, were measured and analyzed.
Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as an SLB with a SGTR.
Development of Device for Detecting Helium Leak From Canister: Part 1 — Experiment on Temperature Behavior During Gas Leak From Canister of 1/4.5 Scale Cask Model
Authors: Takeda Hirofumi , Goto Masanori 
Published: October 2018

A concrete cask has an advantage in cost and period of manufacturing compared with a metal cask. In the metal cask, monitoring of helium pressure between a primary lid and a secondary lid of the cask is required by regulation. On the other hand, in the concrete cask, the lids of a canister are welded and have high sealing performance, so that the monitoring of helium leak from the canister is not required. However, a loss of the sealing performance of the secondary lid, which is caused by stress corrosion cracking (SCC), is concerned in the case of long-term storage by the concrete cask. In the view of this situation, it would be useful to install a helium leak detector in the concrete cask in order to improve the safety of the long-term storage. Thus, we have been developing the detector. The phenomenon that the temperature at the bottom of the canister (TB) increases and the temperature at the top of the canister (TT) decreases during the helium leak from the canister has been confirmed by the experiments in the previous study. We have proposed the performance monitoring by monitoring the temperature difference ΔTBT (= TB-TT) instead of pressure monitoring. This time, to make the installation and maintenance of the helium leak detector easier in consideration of practical use, we proposed a new detection technique using only the temperature of either the lid or the bottom of the canister.
We performed leak tests by using a 1/4.5 scale cask model based on a similarity law of thermal hydraulics. In the experiments, air was used for an inner gas of a canister of the model, and a heat flux of a canister surface had the same value as that of the actual canister surface. In this model, Ra* number can be made to coincide with that of the actual canister.
Besides, Gr* number and Bo* number are almost equal to those of the actual canister. Temperature data at respective canister parts were obtained under condition of canister internal pressure from 6 atm to 1 atm (atmosphere pressure). Also, the new and old leak detection methods were evaluated by using the obtained temperature data.
Study on the Operating Strategy Optimization of Moving to Remote Shutdown Station When Main Control Room Is Un-Inhabitable
Authors: Li Li , Shengtao Zhang , Zhao Xu , Yu Du 
Published: October 2018

For PWR, remote shutdown station (RSS) is a redundant control mean to shut down the reactor when main control room (MCR) inhabitation is challenged (e.g. fire, smoke...). Nowadays, due to nuclear power plants control measures were improved with DCS system, a full function DCS RSS was equipped and more essential equipment could be controlled on RSS.
Under operating conditions that prohibit nuclear power plant operators to stay in the main control room, the operators should move to RSS and shutdown the reactor to ensure plant safety following <Moving to remote shutdown station when main control room is un-inhabitable operating strategy> (RSS strategy for short) to fallback the plant from power operation to cold shutdown. The original operating strategy by nature circulation is no longer the best choice both for operation safety and economy efficiency, and an optimized new strategy should be raised.
Based on the former reason, an optimized operation strategy was raised in this paper. In the optimized strategy, all plant normal standard operation modes were considered as initial conditions, rather than only considering power operation condition in the original one. The fallback mode and fallback strategy for each initial condition was also designed and optimized. To accelerate the depression and heat removal process, a forced circulation operation strategy is adopted when the reactor coolant pumps are available, and less local operation was included by taking advantages of the full function operating measures on RSS. To simplify the whole procedure structure, the operation modules of other general operating procedures are reused.
To validate the effectiveness of the optimized operating strategy, a full scope PWR simulation tool was employed to make thermo hydraulic calculation validation of the reactor response and also the remote control station HMI supporting validation. By simulating the original strategy and the optimized one and related analysis, we found that the optimized strategy is effective, and able to be executed based on the remote control station hardware. By executing the optimized strategy, the unit can fall back to the cold shutdown condition safely and a few hours were saved compared with the original strategy. The optimized strategy had already been implemented on real PWR nuclear power plant.
Two Phase Flow Behavior During Pool Scrubbing
Authors: Nakamura Yuki , Fujiwara Kota , Kikuchi Wataru , Saito Shimpei , Yuasa Tomohisa , Kaneko Akiko , Abe Yutaka 
Published: October 2018

In severe accidents of nuclear power plants, large amounts of gas containing radioactive particles are generated. In the process of gas release into the atmosphere, it is needed to suppress the leakage of radioactive material. The gas is decontaminated by moving radioactive particles from the gas phase to the liquid phase. This effect of capturing particles is called pool scrubbing, and it has been verified great decontamination effect. Therefore, it is extremely important to analyze the effect in evaluating the influence to the environment. But study on its principle is not carried out sufficiently. And also we don’t have enough experimental date to analyze the effect.
The purpose of this study is to clarify the gas-liquid two-phase flow behavior which is important in elucidating the mechanism of pool scrubbing. Particularly, this study focused on the behavior of bubble generation and breakup after being injected from the nozzle and the flow structure of rising bubbles in the still water. Furthermore, we evaluate the validity of the model used in the existing severe accidents analysis code such as the MELCOR by comparing the model with experimental data.
We measure the gas phase jet injected from the upward nozzle inserted to a test water tank. Nozzle diameter, gas phase flow rate, liquid phase temperature, and water depth were used as parameters. Bubble behavior was observed via a high-speed camera. The bubble rising speed, bubble distribution and void fraction were measured by a wire mesh sensor. In previous studies, experiments using non-condensable gas in normal temperature water have been mainly conducted. In order to conduct the experiment under conditions that simulate actual equipment, steam which is a condensable gas was used in this study. Moreover, it is assumed that thermal stratification is formed in the pressure suppression pool during severe accidents. To reproduce this situation, thermal stratification is formed in the test water tank. For bubble behavior and flow phenomenon, the result of using non-condensable gas was compared with that using steam. We consider the influence of formation of a thermal stratification. As described above, the flow phenomenon in the pool scrubbing was visualized and measured. Finally, we discuss the validity of the analysis code by comparing the calculation formula and model in the analysis code with the experiment data.
A Novel More Reliable and Extensible Architecture of Instrumentation and Control Systems
Authors: Zhou Shuqiao , Guo Chao , Li Duo , Huang Xiaojin 
Published: October 2018

Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required.
Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.
The Effect of Functional Spacers on the Liquid Film Thickness and Dryout in a BWR Fuel Bundle Model
Authors: Bolesch Christian , Robers Lukas , Zboray Robert , Prasser Horst-Michael 
Published: October 2018

For the BWRs, the dryout margin is one of the core design limitation factors. Today’s industry standard is to use a large margin to dryout and functional spacer grids with vanes to enhance the heat transfer and to reduce the fraction of entrained droplets. Difficulties for precise measurements under reactor conditions lead to a lack of knowledge on the exact effects of the spacers on the flow and suggest the use of scaled experiments. For this experiment, the goal is to provide high-resolution data for CFD code validation as well as visualizing the effects of functional spacers and the liquid film and potentially the dryout front.
The Dryout Tomography Experiment (DoToX) facility at ETH Zürich is a closed loop experiment for two-phase flow investigations in a fuel bundle model using a modelling fluid. Key aspects are a single undisturbed subchannel and the surrounding four heating rods containing a liquid heating system. This setup allows for a steady state dryout without endangering the structural integrity of the facility and for the 3D reconstruction of the time averaged void distribution within the flow channel by means of an X-Ray and cold neutron Computer Tomography (CT).
In this study we pay special attention to the annular flow in the upper half of the sub channel. We investigate the first results delivered by the facility. Prototypical spacer designs available in the open literature were used. We present the Liquid Film Thickness (LFT) distributions on the walls of the heating rods. Improvements towards the dryout performance as well as drawbacks of the specified spacer design are highlighted.
Heat Transfer to Supercritical Water (Liquid-Like State) Flowing in a Short Vertical Bare Tube With Upward Flow
Authors: Zvorykin A. , Mahdi M. , Popov R. , Barati Far K. , Pioro I. 
Published: October 2018

Current Nuclear Power Plants (NPPs) equipped with water-cooled reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants.
Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV nuclear-reactor concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs.
In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs.
Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required.
Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.
Estimation of External Dose Volume Correction Factors Based on Neural Network
Authors: Chen Junjun , Qu Jingyuan , Gong Junjun 
Published: October 2018

Before the design of the nuclear facility, it is necessary to estimate the dose caused by radioactive material released into the environment during the course of the serious accidents. Semi-infinite hemisphere geometric model is established to estimate the personal external exposure dose outdoors in which the distribution of the nuclides is assumed to be uniform. The exposed staffs are in a limited cubic space when using this model to evaluate the controllability of the main control room. Thus, the volume correction factor is needed to correct the dose, whose traditional expression is f = 352/pow(V, 0.338). The formula cannot satisfy the requirement of higher accuracy due to the neglect of the influence of the shape of geometric model and γ-rays energy. Usually the actual control room is a cube and the γ-rays energies emitted from various nuclides are different.
In order to calculate the accurate volume correction factor of main control room under different geometric conditions, a finite cubic geometric model is established in this paper. The length and width of the model are between 6m and 50 m, the height is between 4m and 6m, and γ-ray energy respectively are 0.05, 0.2, 0.733, 1.2 and 3 MeV, respectively. The effective volume values for different conditions are calculated by the Monte-Carlo program, and 318 groups of results are obtained. The calculated volume dose rate of 360m × 360m × 255m (assuming semi-infinite) cube at 733keV γ-rays energy is taken as a criterion, whose ratio of the other calculation results is the new volume correction factor value. By comparing two volume correction factors, the relative discrepancies are within 3 folds, proving that the calculation result is reasonable and feasible. The new volume correction factor varies with γ-rays energy and the shape of the geometric model.
A neural network model corresponding to the volume correction factor is developed to apply to more cases. 80% of the results are randomly selected as the training set of neural network. The remaining 20% of the result as the test set of the cross-test is to predict the results of the trained neural network, whose relative errors are less than 5%. The neural network model can obtain the volume correction factor under different geometry and γ-ray energy conditions. Finally, a volume correction factor library is established, which can provide a powerful reference to obtain the volume correction factor of the limited space model such as the main control room.
Research on Thermal Efficiencies of Various Power Cycles for GFRs and VHTRs
Authors: Mahdi Mohammed , Popov Roman , Pioro Igor 
Published: October 2018

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs.
Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle.
Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N2 mixture, application of SuperCritical (SC)-CO2 cycles or combined cycles.
The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO2 is provided.