bluuebunny/crossref_metadata_2025_split huggingface dataset:
Heat Transfer Simulation of the Fuel Transport Cask for Spherical Fuel Elements in Molten Salt Reactor
Authors: Liu Yang , Wang Jun 
Published: October 2016

Fuel transport is an indispensable task for nuclear power plants. For pressurized water reactors (PWR) and boiling water reactors (BWR), many research projects have been completed for designing and testing the transport casks for fresh fuel as well as spent fuel [1–3]. To ensure the safety of nuclear fuel during the transportation, many aspects should be analyzed and examined for the casks with fuel inside, such as heat transfer and temperature calculation, radiation protection, nonproliferation issues, etc. The transport cask discussed in this paper is especially for new spherical fuel elements, which should be designed in accordance with the stipulations in the GB11806 Regulations for the Safe Transport of Radioactive Material [4].
The Transport Cask for spherical fuel elements used in molten salt reactor (MSR) should be designed in accordance with the safety standards for transport of radioactive material. It is necessary to evaluate the thermal performance of the transport cask separately in normal transport condition and in accident transient. The MSR fuel sphere elements cask is in a circular cylinder shape and composed of inner container and outer shell cask. The objective of the thermal analysis of the cask under hypothetical accident conditions is to demonstrate that the cask containment boundary structural components are maintained within their safe operating temperature ranges. The heat transfer process (conduction, convection, and radiation) is simulated by ANSYS-CFX in this paper and it is demonstrated that the components of cask are maintained in safe operating temperature ranges. The calculation results are below limit temperatures, indicating that the thermal design of the cask could meet the Standard Regulations. The result is also compared with the fire test, which shows the calculation model is conservative and rational.
Assessment on an Intelligent Controller of Marine Condenser
Authors: Xian Gong , Jianhua Sun , Hai Zou 
Published: October 2016

At present, simulation research on condenser mostly focuses on the one of power plant. While little attention is paid on the marine one. The marine condenser is different from the one of power plant. One difference between them is that marine condenser limited by the hull space is much smaller than the one of power plant. Another difference is that marine condenser usually runs in condition that steam load changes greatly and frequently. In addition, the temperature of cooling water changes with seasons and maritime space. When the ship makes an ocean-going voyage, the temperature of cooling water varies obviously, which has a negative effect on the control system. Therefore, simulation research to the dynamic characteristic of marine condenser is carried on by this paper.
The vacuum and the condensate supercooling degree are the main operating parameters for a condenser. In this paper, a mathematical model for marine condenser is established. The dynamic variation of each parameter is monitored by this model. Then an expert control system is designed. The condenser vacuum and supercooling degree are controlled by the flow rates of cooling water, air-blend and steam jet for deaerating. The deaerating steam bubbles which are not absorbed completely by condensate water, have a negative impaction on the condenser vacuum. The deaerating steam bubbles heating the condensation water to be exactly saturated is calculated. Referring to it, the fluctuation of condenser vacuum is filtered. So the negative influence on condenser control system can be remitted. The temperature of water replenishing is usually lower than that of condensate water in condenser heat well. And the water is replenished into the heat well part. So water replenishing would increase the supercooling degree of condensate water.
This paper adopts the simulation experiment method to gain the expert knowledge of condenser control. According to power level, inlet temperature of cooling water, the messages of operating modes and sensors, the control system regulates the regulating valves for turbo circulating pump of cooling water, air extractor and steam jet deaerator to change the flow rates of cooling water, air-blend and steam jet for deaerating, so as to achieve automatic control of condenser vacuum and supercooling degree. Finally, two PID controllers are added as well, to counteract the adverse effect, enhance anti-interference ability and optimize control performance of intelligent control system. Simulations of varying load and cooling water temperature are done. Results show that the intelligent control system could effectively control condenser vacuum and supercooling degree under different power level and variable temperature of cooling water condition.
3D Linear Bifurcation Analysis of Steel Containment Vessel
Authors: Schmidt Jay , Laskewitz Bernd 
Published: October 2016

The AP1000® Containment Vessel (CV) is a freestanding steel containment designed to protect the public from radiation release. The CV consists of 2 ellipsoidal heads connected by a cylindrical shell and is constructed of carbon steel. The AP1000 plant design has four large penetrations (two airlocks and two equipment hatches) located in approximately the same quadrant of the circumference of the shell which imposes asymmetric effects in the shell structure. The CV is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Subsection NE.
Traditionally, the local and global stability of freestanding steel containments have been designed by use of formulae using conservative assumptions based on an axisymmetric structure. ASME Code Case N-284 “Metal Containment Shell Buckling Design Methods, Class MC Section III, Division 1” outlines methodology for satisfying the stability of the CV using two approaches. Section 1710 provides a stress based buckling approach using detailed formulae that assumes an axisymmetric structure. The second approach provides guidance and acceptability based on a linear bifurcation analysis (2D (1720) or 3D (1730)). Due to the asymmetric structure of the CV, the 3D linear bifurcation method delivers the most accurate results.
The methodology and assumptions implemented by Westinghouse to qualify stability of the CV via Code Case N-284 are outlined. Also, the procedure to properly amplify the stresses as required by N-284 is included as justification of the methods used. This justification was thoroughly investigated by the Nuclear Regulatory Commission (NRC) and deemed acceptable.
The Comparison of the Transducer Circuits for the Two-Phase Flow Conductivity Probe
Authors: Yao Zhipeng , Ge Liangjun , Abbas Kamil , Shiliang Zhou 
Published: October 2016

The void fraction of two-phase flow can be measured by conductivity probe due to the difference of conductivity between air and water. It’s difficult to detect the conductivity of the probe directly so transducer circuits should be used to transfer the conductivity to voltage, and the voltage can be sampled by DAQ device. Two transducer circuits are designed for two-phase flow double sensor conductivity probe. One is a simple DC circuit, it’s made up of a constant DC voltage and a resistance, and the resistance is cascaded with the probe. Voltage drop across the resistance is depend on conductivity of surrounding media of probe tip. The other is AC circuit that utilizes a DDS (Direct Digital Synthesizer) as an excitation power supply which links to a transimpedance amplifier through the probe. The current that flow to the amplifier would change according to the surrounding’s medium of the probe. The transimpedance amplifier transfers the current to voltage. Then we use AD8302, an amplitude detection chip, to detect the amplitude of the s output.
The experiments show that the DC circuit has a higher SNC (signal to noise ratio) and shorter responding time. However, the needle of probe is covered with electrochemical corrosion products and its performance degrades significantly when DC circuit is used and the probe is immersed in 80°C hot water for 8 hours. This kind of phenomenon is not appeared on the probe when DDS-based circuit is used. The result of the experiment verifies the SNC and responding time of DDS-based circuit is approximately equal to that of the DC circuit.
Investigation on Thermal Treatments of High-Resistance Austenitic Stainless Steels Employed as Structural Materials in Generation IV Reactors
Authors: Mazzei Giulia B. , Lo Frano Rosa , Stubbins James F. 
Published: October 2016

Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr–25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure.
This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.
Cryogenic Hydrogen Cooling of Heated Moderator Vessel
Authors: Nilsson Per , Pucilowski Mateusz , Bessler Yannick 
Published: October 2016

In ESS, a pulsed proton beam of 5 MW mean power will hit a tungsten target to generate neutrons by spallation. The pulses are 2.86 ms long and occur with 14 Hz; the power within a pulse is 125 MW. Only centimeters from the target, the neutrons are moderated by liquid hydrogen in aluminium vessels.
The deposited power heats the surrounding structures and fluids which are circulated and cooled. The hydrogen is operating at 15 bar and average temperature between 17 and 21 K, i.e. above the critical pressure 12.8 bar, but below the critical temperature 32.9 K. During the pulses, the peak heat deposition in the aluminium is 15 W/cm3 and in the hydrogen 4 W/cm3.
If the cooling of the aluminium is neglected during one pulse, the temperature increases to 34 K. That is above the critical temperature, where physical properties change strongly with temperature. Therefore the conjugate heat transfer has to be investigated in detail.
This work includes 1D principal transient calculations of a general configuration as well as CFD simulations of the heating and cooling of a specific design. The 1D calculations are performed using GNU/Octave and the CFD using ANSYS/CFX. It is concluded that with an inlet temperature of 17 K, the wall temperature can be kept below the critical temperature in the general configuration and sufficient cooling can be ensured in the investigated specific design.
Impact Resistant Design for FLEX Building Equipment Door
Authors: Haider Shakir Y. , Calhoun David J. 
Published: October 2016

After the 2011 Japan earthquake and tsunami caused significant damage at Fukushima Daiichi, the Nuclear Regulatory Commission required all US nuclear power plants to have a mitigation strategy for beyond design basis events. Industry-developed response plans, called “FLEX” strategies, deployed new, portable equipment such as diesel generators and cooling pumps. As this new equipment needs to be available after a natural disaster, storage in protected locations is required. Many nuclear plants have recently constructed new storage buildings, or FLEX buildings, as part of their post-Fukushima strategy.
The equipment door is a critical component of a FLEX building. Large enough to drive a semi-trailer truck through, it is required to protect the equipment in case of an earthquake, flood, tornado and also may need to be capable of opening within a few minutes in order to respond during an emergency. The equipment door presented in this paper serves these purposes very effectively. The composite section of the door is capable of shielding the structure from penetration as well as overall dynamic response from tornado missile impact. The door travels on an overhead rail which, being indoors and above the opening, provides reliable door movement in case of snow or ice during winters or in case of debris from wind or tornado. Latches capable of withstanding tornado missile impact forces also restrain the door in case of wind or seismic forces. The door is opened by means of motorized trollies and is also equipped with a backup opening device by means of an air winch in case of a power loss.
The door and the latches that restrain the door from impact are analyzed using ANSYS finite element software. A limit state analysis is performed that identifies the sequence of yield limit states for the components of the door and the door latch as the loading progresses. The analysis continues until the ductility limit state for the latch is reached. Redistribution of stresses within the components of the latch is observed during the analysis. A modal analysis and a direct integration time history analysis is also performed to capture the dynamic response due to impact loading.
Overall, this paper presents a highly robust and reliable design for a FLEX building equipment door that is capable of protecting the contents of the building during a natural disaster and remaining operational during the response after an emergency.
Transient Heat Transfer for Helium Gas Flowing Over a Horizontal Cylinder Under Flow Decay Conditions
Authors: Liu Qiusheng , Wang Li , Shibahara Makoto , Fukuda Katsuya 
Published: October 2016

The Very High Temperature Reactor (VHTR) is a new reactor that uses helium gas as primary coolant for conventional graphite matrix, coated fuels. It enables the achievements of high thermal efficiency and can supply heat with a high temperature of about 900–1000 °C. During the loss of coolant process, the fuel hot spot temperature should not get over a criteria value due to the temperature limitation of the fuel assembly. Traditionally, the VHTRs are designed to deal with loss of forced circulation conditions by using a passive mode decay heat removal system for the cavity cooling. However, even passive systems may experience some failure due to multiple undesired conditions even though the possibility is extremely low. Therefore, the VHTRs are now expected to be designed as naturally safe reactors with inherent safety features. Which means the decay heat removal is fully dependent on natural convection and radiation. To accomplish the tough task, a clear understanding of the heat transfer process during flow decay transient condition is quite necessary.
This study was conducted to investigate the transient heat transfer process between the solid surface and coolant (helium gas) in VHTR under flow decay conditions. Forced convection transient heat transfer for a horizontal cylinder under flow decay transient condition was experimentally studied. The experiment was conducted by using the helium gas as coolant. A uniform heat generation rate was added to the heater. With a certain flow rate of the helium gas, the heater temperature was maintained at a designed value. Then, the flow rate of the helium gas starts to decrease according to designed linear functions with different decreasing speed. Platinum cylinder with 1 mm in diameter was used as the test heater. The heat transfer coefficient and surface temperature were measured during the flow decay transient process under wide experimental conditions such as initial flow rate, flow decay time. It was found that the temperature of the test heater increases in curve shape with different gradients during this process, with a shorter flow decay time the increasing rate of heater surface temperature would be higher. The heat transfer coefficient versus time during the flow rate decreasing process was also obtained.
Sorption of Radioactive Methyl Iodide by Silver Doped Zeolite for Filtered Venting System
Authors: Ishii Tasuku , Akiyama Daisuke , Kirishima Akira , Sato Nobuaki , Narabayashi Tadashi 
Published: October 2016

Filtered containment venting system (FCVS) is an emergency response system for severe accident. FCVS reduces the pressure in reactor vessel and removes radioactive iodine and cesium.
Recently, silver doped zeolite (AgX) has been attracting attentions since it is believed to remove over 99.99% of CH3I in the ventilation gas. However, the reliable data of AgX for CH3I sorption under expected severe accident conditions are still limited. Therefore, the sorption behavior of CH3I on AgX was investigated under several expected conditions.
In this study, the CH3I labeled by 125I tracer was synthesized by isotope exchange reaction, and it was devoted for sorption experiment on AgX. Sorption experiment using stable CH3I was also conducted by the same procedure. From results of the sorption experiment, over 99.9% of 2.28g of CH3I were sorbed by 10g of AgX. After the sorption experiment, AgX samples were analyzed by powder X-ray diffraction (XRD) and SEM–EDX. These results indicated the distribution of Ag agreed with that of I in micro area of AgX. For the evaluation of apparent sorption capacity of AgX, breaking through experiment was carried out. The apparent sorption capacity of AgX for CH3I was determined as 0.25g / g (AgX) at 24°C. The breaking through experiment was also carried out for the evaluation of CH3I flux and humidity effects in the vent gas. The effect of temperature on the sorption capacity was also studied. When the flux of CH3I supply was increased from 1.09 × 10−4 to 2.33 × 10−4g/cm2s, apparent sorption capacity was unchanged (= 0.21g / g (AgX)). Besides, the apparent sorption capacity was unchanged in the humidity range of ∼0 % H2O to 0.01 % H2O, while it decreased to 0.19 g / g (AgX) at 3% H2O. When sorption temperature increased to 170 °C, the apparent sorption capacity also increased to 0.69 g / g (AgX).
Development of “Proactive Safety Review Guideline for Continuous Improvement on Nuclear Power Plants”
Authors: Narumiya Yoshiyuki , Okamoto Koji , Ueno Shingo , Hayashi Kentaro , Kuramoto Takahiro 
Published: October 2016

The Atomic Energy Society of Japan (AESJ) has established and issued the “Proactive Safety Review (PSR+) Guideline for Continuous Improvement on Nuclear Power Plants: 2015” 1) through the discussion at the Periodic Safety Review Subcommittee under the System Safety Technical Committee of the AESJ Standards Committee (SC), and then approved by the both committees. This guideline defines the review concept, perspectives, and methodologies, relating to identifying areas of safety improvement when the licensees implement periodic review for continuous improvement of safety of their nuclear power plants, so called “Proactive Safety Review (PSR+).
This paper discusses the details of this newly established standard, including the concept, perspectives, and methodologies, relating to identifying areas of safety improvement. In addition, specific examples of applying this standard and the policy of utilizing this standard in continuously improving the safety of all the nuclear power plants in Japan are also described.
Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident
Authors: Zanchetti Alexandre , Hassanaly Mickael , Cordier Hervé , Sanna Antonio , Mechitoua Namane , Mimouni Stéphane 
Published: October 2016

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations.
The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.