bluuebunny/crossref_metadata_2025 huggingface dataset:
The Effects of a Non-Condensable Gas on Pressurizer Insurge Transients Under the High Pressure
Authors: Bolong Wang , Weihua Li , Haijun Jia , Jun Li
Published: October 2018

The prediction of the pressurizer response during various transients plays an important role in determining what actions should be taken to control or mitigate the consequence of an accident. For this reason, a range of different experiments have been completed which aim to support a preliminary understanding of the phenomena which are important to the analysis of a PWR pressurizer.

As so far, the researches on these subjects are focusing on the low pressure, there are rarely researches on the high pressure. Hence, the effects of a non-condensable gas on pressurizer insurge transients under high pressure are necessary and significant.

The study concentrates on the pressure response and heat transfer during an insurge transient under the high pressure. These experiments were run in two cylindrical stainless steel tanks: the primary tank and the storage tank. In this research, the experiments looked at the effects of water level increase and the presence of a non-condensable gas with varying mass fraction. Nitrogen was chosen as the non-condensable gas. The storage tank was filled with water and pressurized with nitrogen. An insurge to the primary tank with hot walls was initiated by opening the high pressure solenoid valve between two tanks. The experiments assigned lots of thermocouples for monitoring the radial and axial temperature distribution. Experiments showed that the non-condensable gas has a substantial influence on the pressure response. And the magnitude of the increase in peak pressure is quite different with the change of experiment initial conditions.
CFD Modelling of Flashing Instability in Natural Circulation Cooling Systems
Authors: Liao Yixiang , Schuster Christoph , Hu Suqing , Lucas Dirk
Published: October 2018

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.
CFD Numerical Simulation of Water Hammer in a Vortex Diode
Authors: Wang Junrong , Wei Zhiguo , Gou Jinlan , Xiao Qi , Li Shaodan , Li Yong
Published: October 2018

In the pipeline system of nuclear industry, shock wave pressure in a pipe will be caused by the fast closing check valve after the pumping stops. This phenomenon is known as water hammer, which brings hidden danger to the security and reliability of the pipeline system. Specially, water hammer may cause serious damage on the pipeline system by the valve misoperation, by the valve malfunction, or by other unexpected events. A vortex diode is used as a highly reliable check-valve in nuclear applications, where it mainly benefits from the intrinsic properties of no moving parts and no leakage. In this paper, we proposed a novel method based on a vortex diode to protect water hammer.

In the traditional analysis, a simple one-dimensional (1D) model is often used to simulate the water hammer. However, it is difficult to get the transient flow characteristics in a vortex diode using a 1D model. Thus, a three-dimensional (3D) model using computational fluid dynamics (CFD) is proposed to analyze water hammer in a pipeline system with a vortex diode. The 3D model was firstly verified by comparing the numerical results of CFD with experimental results of a water hammer test. Based on the 3D model, the water hammer was simulated at different inlet conditions in a pipeline system with a vertex diode. In order to investigate the vortex diode used as a leaky check-valve, the inlet pressure was decreased by the corresponding value of pump head to simulate the pump stop after the quasi-steady state was achieved in the vortex diode. It is found that the pressure fluctuation of water hammer is comparable to the pump value, which is not varying with initial velocity in the pipeline system. Thus, we have proved that a vortex diode in the pipeline system acts significantly in suppressing pressure fluctuation of water hammer. This study presents a CFD-based numerical method for water hammer and could be useful in protecting water hammer in nuclear industry.
A Numerical Study of Particle Deposition Through Fuel Pebble Bed in HTGR
Authors: Sun Qi , Zhao Gang , Peng Wei , Yu Suyuan
Published: October 2018

The study on the deposition of graphite dust is significant to the safety of High-Temperature Gas-cooled Reactor (HTGR) due to potential accident such as localized hot-spots and intensity change which is caused by the graphite dust generated by abrasion of fuel elements. Based on the steady flow and three-dimensional face centered structures of fuel pebble bed, the discrete phase model (DPM) were applied to simulate trajectory of graphite dust in conditions of HTGR. To determinate the deposition of particle, the present study introduces a rebound condition with critical velocity by a user defined function. The particle trajectories show most of particle deposition can be summed up as the effect of backflow region, turbulent diffusion and inertial impact. The original trap condition overestimates the deposition fraction especially for large particles compared with involving rebound condition. In addition, the trend of deposition fraction shows as the dimeter of particle increases, deposition fraction decreases first and then increases.
Expected Accuracy Range of Cost Estimates for Small Modular Reactors at the Early Concept Design Stage
Authors: Agar Amritpal S. , Fry Andy J. , Goodfellow Martin J. , Goh Yee M. , Newnes Linda B.
Published: October 2018

Life cycle cost is an important consideration for the development and selection of new power generation technology. Large nuclear power plants (NPPs) have been subject to capital cost escalation, stemming from delays related to late design changes, procurement issues for major components, and regulatory enforced changes. These factors have contributed to the significant risk premium associated with gigawatt scale “Gen III+” designs, which have incurred significant financing costs. Large NPPs have become prohibitively expensive for many utility investors in liberalized markets and smaller economies.

The challenge of reducing upfront capital costs is one of the requirements that have driven the development of innovative Small Modular Reactors (SMRs). These designs are said to offer reduced unit cost and reduced risk due to certainty of delivery, which could lead to a lower cost of capital for a utility customer. By offering a product with more cost certainty the SMR could restore investor confidence in nuclear power. The life cycle cost estimates associated with the different SMR designs are uncertain at the early stage of development. However, designers need to understand, with some confidence, the impact of technical decisions at the early development phase on the life cycle cost.

This study presents an overview of cost uncertainty associated with the early design stage of the SMR. The types of cost estimating approaches available at the concept design phase are identified and categorized in terms of their expected accuracy ranges. The Overnight Cost of Construction (OCC) is an important driver of the life cycle cost of a power generation project. The expected accuracy ranges from each estimating method are used to illustrate the sensitivity of cost uncertainty to the level of design maturity. By understanding the sources and impact of cost uncertainty decision making during product development can be optimized to meet both technical and commercial requirements.
Inlet Passageway Optimization of Immediate Heat Exchanger in an HTGR
Authors: Cui Jingdan , Yuan Kun , Sun Qi , Peng Wei , Wang Jie
Published: October 2018

The immediate heat exchanger is important for the heat utilization and hydrogen production in a very high-temperature gas-cooled reactor (VHTR). There is a 90-degree elbow passageway in the inlet passageway of immediate heat exchanger, contributing uneven flow distribution in circumference direction in the heat transfer area, which is harmful to the heat transfer tubes aging process. In this study, three inlet passageway optimization designs are put forward and their optimization effect is analyzed according to specific criterion by CFD. As the existence of spiral tube area would have coupling effect on inlet passageway flow, in the computation, the heat transfer area is assumed to be porous medium, in which the effect of heat transfer is taken into account as well. The result shows that installing flow distributer in the inlet passageway could make helium flow on circumference direction most homogenous.
Numerical Study of Supersonic Film Cooling in Diverging Section of Nuclear Rocket Laval Nozzle
Authors: Sun Xiaokai , Ye Ping , Jiang Peixue , Peng Wei , Wang Jie
Published: October 2018

Nuclear rockets with specific impulse have obvious advantages by greatly reducing the mass of the propellant and potentially decreasing the cost of launching material from the earth’s surface. Nuclear thermal rockets use hydrogen propellant with coolant exit temperature of near 3000 K, which is very high, so the cooling of airframe surfaces in the vicinity of the exhaust is needed, of which film cooling is an effective method. Most of previous studies mainly focus on the film cooling effectiveness using two dimensional backward-facing step model, however, the nuclear rocket exhaust using the converging-diverging Laval nozzle, so the film cooling would be different. The present study numerically investigated the influence of coolant Mach number, coolant inlet height on supersonic film cooling in the diverging section of Laval nozzle, while keeping the coolant mass flow rate constant, with the results showing that: increasing the coolant inlet Mach number and the coolant inlet height can increase the film cooling effectiveness; for the same coolant mass flow rate, reducing the coolant inlet height and increasing the inlet Mach number improves film cooling effectiveness.
The Cloud Model Theory of Intelligent Control Method for Non-Minimum-Phase and Non-Self-Balancing System in Nuclear Power
Authors: Tao Mo , Qu Ruotong , Ke Zhiwu , Chen Zhaoxu , Li Xianling , Feng Yi
Published: October 2018

Steam generator (SG) is one of the key equipment of nuclear power units. Because of the large range of its loads changing, the water level control of SG effectively is an essential secure guarantee of nuclear power plants. SG is a complex system, besides imbalance and non-minimum phase characteristic, it also has the properties of nonlinearity, time-varying and with small stability margin. There are many difficulties in water level control of SG. Of which false water level and varying parameters are the most severe problems.

In this paper, first the water level features and the water level control principle of U-tube steam generator (UTSG) are introduced. Then mathematical model mechanism and both the static and dynamic characteristic of it water level are discussed. Finally various control methods are used for comparing the control effect.

Intelligent control is a type of control strategy which imitates human intelligence behavior. It is mainly aimed at the controlled plant with complicate model parameters, or which model structure hard to describe accurately by mathematical method. Cloud Model theory is proposed by Academician Li Deyi based on the idea of artificial intelligence with uncertainty. This theory focus on analyzing the uncertainty of control plant, realizes the uncertain conversion between qualitative concept and quantitative numerical by combining ambiguity and randomness. In the field of control technology, ambiguity and randomness make it difficult to establishing precise mathematical model of control plant, and become a bottleneck during the research of improving stability, accuracy and quickness of control system.

In this context, Cloud Model can be a good conversion between qualitative concept and quantitative numerical due to its ability of showing the uncertainty of qualitative concept which described by natural language. Under the action of external input, system control can be realized by inferencing according to the qualitative concept and uncertainty rules of Cloud Model. In this paper, the researched Cloud Model control system is based on normal distribution, because a large number of random events in nature and society obey or approximately obey normal distribution.

The rate of convergence of Cloud Model control is evidently faster than PID. Moreover, the capability of Cloud Model control in tracking, adapting, anti-interference and overcoming large time lag are apparently superior when comparing with the control effect of PID.
Development of Laser Cleaning for Metallic Equipment
Authors: Carvalho L. , Pacquentin W. , Tabarant M. , Lambert J. , Semerok A. , Maskrot H.
Published: October 2018

Laser cleaning study was performed on prepared samples using a nanosecond pulsed ytterbium fiber laser. To prepare samples, AISI 304L stainless steel samples were oxidized and implemented with non-radioactive contaminants in a controlled manner. In order to validate the cleaning process for metallic equipment polluted in nuclear installations, two types of contamination with europium (Eu) and with cobalt (Co) were studied. Eu was used as a simulator-product resulting from uranium fission, while Co — as an activation-product of nickel, which is a composing element of a primary coolant system of a reactor. The oxide layers have suffered laser scanning which was followed by the furnace treatment to obtain thicknesses in the range of 100 nm to 1 μm depending on the oxidation parameters [1] with a mean weight percentage of 1% of Eu and 1 % of Co in the volume of the oxide layer. Glow Discharge Optical Emission (GD-OES) and Mass Spectrometry (GD-MS) analyses have been performed to assess the efficiency of the cleaning treatment and to follow the distribution of residual contamination with a detection limit of 0.1mg/kg of Eu and Co. Decontamination rates up to 95.5 % were obtained.

One of the identified reasons for this limitation is that the radionuclides are trapped in surface defects like micro cracks [2, 3]. Therefore, cleaning treatments have been applied on surface defects with controlled geometry and a micrometric aperture obtained by laser engraving and juxtaposition of polished sheets of AISI 304L stainless steel. The goal of this study is surface decontamination without either welding or inducing penetration of contamination into the cracks. GD-MS analysis and Scanning Electron Microscopy (SEM) were performed to analyze the efficiency of the treatment and the diffusion of contaminants in this complex geometry.
A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship
Authors: Yan Shiyu , Liu Hua , Liu Zhaohui , Yang Xiaohua , Li Meng , Chen Zhi
Published: October 2018

In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.
The Design of Data Structure and Interface for Nuclear Emergency Assessment and Decision Support System
Authors: Yapeng Yang , Hong Wei , Zongyang Feng , Linsheng Jia , Xiaoxiao Xu , Jiangang Zhang
Published: October 2018

Emergency Assessment and Decision Support System has become an important tool to effectively respond to a nuclear emergency in nuclear power plant, the system mainly used to on-line assessment the real-time reactor conditions, calculate the source term released to the environment, and suggest the protection actions taken to protect the public and emergency workers. This paper describe the development of the system, analysis the business flow chart of nuclear assessment, introduce details of several functional modules, such as core damage assessment, source term calculation and radiological consequence assessment. The system’s data structure and data interface also be describe in detail.
AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model
Authors: Estevez-Albuja Samanta , Jimenez Gonzalo , Fernández-Cosials Kevin , Queral César , Goñi Zuriñe
Published: October 2018

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results.

In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.
AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants
Authors: Gautier Bernard , Cesbron Mickael , Tulinski Richard
Published: October 2018

Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.
The Backfit Rule’s Compliance Exception
Authors: Miranda Samuel
Published: October 2018

The Backfit Rule, 10 CFR §50.109, requires the NRC staff to produce cost-benefit evaluations to justify any changes it may make in its positions, or any new requirements it might impose on licensees, for the purpose of enhancing plant safety. The Backfit Rule also allows the NRC staff to forgo cost-benefit evaluations when it identifies errors or omissions in licensing submittals, or in its own reviews of licensing submittals. In such cases, there is no safety enhancement to be realized. Instead, the NRC staff seeks to obtain reasonable assurance that the level of safety, required by regulations, and licensees’ commitments, is maintained or, if necessary, restored. This provision in the Backfit Rule, called the “Compliance Exception”, has been proven to be very difficult to apply.

In 1998 and in 2001 two licensees had argued, in License Amendment Requests (LARs) that their pressurizer safety valves (PSVs) were qualified for water relief duty. Consequently, their PSVs were safety grade components that could be assumed to be available, in licensing basis accident analyses, to open, relieve water, and then reseat. This capability was thought to be required in order to mitigate certain accidents that caused the pressurizer to become water-solid. The 1998 LAR was withdrawn when the licensee was informed that its PSV test results did not demonstrate a capability to relieve water. However, the 2001 LAR was approved, based upon the licensee’s claim that it had acceptable PSV test results. Later, in 2013, the NRC staff realized that the PSV tests, cited in the 2001 application, did not actually exist. So, after careful consideration, over a two-year period, the NRC staff issued a compliance-based backfit order to the licensee. The licensee appealed the order, and the NRC staff denied the appeal. Then the licensee filed a second appeal, this time directly with the NRC’s Executive Director of Operations (EDO). (Such appeals are allowed by the Backfit Rule.) The EDO granted this appeal. So, the compliance-based backfit order, which was intended to address the missing PSV test results, was ultimately overturned. The PSV test results are still missing; but the licensee now has the NRC’s approval to assume the operation of water-qualified PSVs in its licensing basis accident analyses.

This paper follows the writing, issuance, and appeal of this compliance-based backfit order, and describes how difficult it is to apply the Compliance Exception of the Backfit Rule.
Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes
Authors: Horie Hideki , Takeuchi Yutaka , Takiwaki Kenya , Sebe Fumie , Kakiuchi Kazuo , Sato Hisaki
Published: October 2018

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry.

Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction.

In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case.

On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.
Incorporation of the Radioactive Interinfluence in Multi-Unit Seismic PRA
Authors: Matsunaka Shuhei , Sato Chikahiro , Watanabe Manabu
Published: October 2018

Kashiwazaki-Kariwa nuclear power station of TEPCO is the largest nuclear power station in the world, and it has seven nuclear power plants. As the experience at Fukushima Daiichi nuclear power station accident in March 2011 involving concurrent core damage at multiple units, it is considered that the risk derived from hazards of Earthquakes and Tsunamis is relatively significant in Japan, and these events have a high likelihood of damaging multiple units simultaneously. Therefore, it is very important to grasp the multi-unit specific risk.

Although there are some unique accident scenarios of Multi-Unit PRA, this paper focuses on the influence of radioactive materials released outside the containment vessel on the accident management of the adjacent unit. The events including core damage and loss of containment function should be considered as the causes of the release of radioactive substances, and operator’s operation or the like should be considered as objects to be adversely affected by them. It is necessary to incorporate that into PRA to confirm the effect on risk.

It is very difficult in terms of the maturity of evaluation method and the calculation load to accurately incorporate consequences derived from time series of various events and complicated interaction into PRA model. Therefore, as the first step in evaluating the risk of influence of radioactive material release on the accident management, some streamlining efforts are implemented according to the purpose. For example, Kashiwazaki-Kariwa unit 6 and unit 7 were set as the target units for model simplification. We also assume the earthquake as the initiating event due to the strong common factor for multi units. Whether or not to be operable in the adjacent plant is set conservatively based on deterministic evaluation.

PRA taking into consideration the radiation influence by multi-unit accident is compared with normal PRA. Some kind of Core Damage Frequency (CDF) such as CDF1 (Core Damage Frequency at which the damage of one or the other of two unit occur), CDF2 (CDF at which the damage of both of units occur) and CDFTOTAL (CDF at which the damage of one or more units occur: CDF1 + CDF2) are quantified, and the degree of this issue is provided. Although the change of CDFTOTAL was insignificant, the necessity of further study was shown from the viewpoint of the amount and timing of radioactive substance released due to an approximately 1.5-fold increase in CDF2.
Study on the Mechanism and Characteristics of Transient Noise of the Steam Discharge Pipes
Authors: Dai Lu , Wei Zhiguo , Wu Jun , Liu Yong , Ma Can , Xiao Qi
Published: October 2018

While the steam power plant runs at low load, the drastic vibration and noise of the steam discharge pipes will be induced by the discharging process of high temperature and high pressure steam. Motived by this, the analytical model of inner flow field of the steam discharge pipes is modeled in this paper. The large eddy simulation method is used to analyze the characteristics of the inner flow field of the steam discharge pipes. The mechanism of transient noise in the process of steam discharging, by the comparison of variation law of the inner flow field with different opening angles of control valve or different time, is obtained. On this basis, the K-FWH method is employed to calculate the transient noise of steam discharging. The results show that the key factors of transient noise generation can be owed to the impact effect of steam jet flow on the surface of valve structure, the shear and friction effect induced by the uneven flow, the energy loss of vortex collapse in the process of valve throttling, and the surface noise sources induced by turbulent flow and so on. The throttle noise induced by control valve is presented as significant broadband characteristics in the frequency domain, and the level of transient noise increases with the decreasing of the opening angle of control valve.
LBLOCA Initiated Emergency Condition Analysis for a China Three-Loop PWR
Authors: Ning Wang , Lei Chen , Jiangang Zhang , Yapeng Yang , Xiaoxiao Xu , Zongyang Feng , Linsheng Jia
Published: October 2018

Great interest in severe accident has been motivated since Fukushima accident, which indicates that the probability of severe accident exists even though it is extremely small. Emergency condition is important in decision making in case of severe accident in NPP. Although many studies have been conducted for severe accident, there was necessary to investigate emergency condition of severe accidents that could possibly happen and haven’t been sufficiently analyzed. Since station blackout (SBO) happened in Fukushima accident, a number of studies in severe accidents initiated by SBO have been carried out. Off-site power is assumed to be lost during large break loss of coolant accident (LBLOCA), but there is few study to find out emergency condition during LBLOCA if both of off-site and on-site power are lost. A hypothetical severe accident initiated by LBLOCA along with SBO in a China three-loop PWR was simulated in the paper using MELCOR code. Emergency condition was obtained including start of core uncover, start of zirconium-water reaction, failure of fuel cladding and failure of the lower head. Thermal-hydraulic response of the core during the accident was also analyzed in the paper.

The model for this study consists of 46 control volumes (27 in primary loop, 17 in secondary loop, 1 in containment and 1 in environment) and 52 flow paths. High pressure safety injection (HPSI) and low pressure safety injection (LPSI) are lost because of loss of on-site and off-site power, and simultaneously main feed water and auxiliary feed water of the steam generators are lost for the same reason. The accumulator can inject water into the core since it is passive and doesn’t need any power.

Results of the study will be useful in gaining an insight into detailed severe accident emergency condition that could happen in a China three-loop PWR and may provide basis for severe accident mitigation.
CFD Modelling of Loss of Vacuum Accident (LOVA) for CFETR
Authors: Xu Youyou , Liu Songlin , Cheng Xiaoman , Ma Xuebin
Published: October 2018

Chinese Fusion Engineering Testing Reactor (CFETR) is a tokamak-type machine and next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the International Thermonuclear Experimental Reactor (ITER) and the demonstration reactor (DEMO) [1]. The accident sequence starting from loss of vacuum accident (LOVA) is an important issue concerning the performance of CFETR. During LOVA, air will leak into the vacuum vessel (VV) causing fast pressurization of VV. At the same time, the high speed airflow jet will result in migration and re-suspension of the large quantity of tungsten dust produced and deposited in the lower part of plasma chamber, causing possibilities of radioactive dust leakage into the workshop and environment. In order to conduct a comprehensive analysis of the accident sequence, firstly, the airflow characteristics of LOVA should be studied. In this article, a postulated rupture of different section area is assumed due to a failed component at the equatorial port level. The computational fluid dynamic (CFD) modelling of LOVA was conducted by ANSYS CFX. The results show that the break area has significant influences on the characteristics of the airflow. Two swirling airflows are formed in the upper and lower part of the torus. The airflow characteristics are quite different when the LOVA happens during maintenance or during normal operation. A reverse flow occurs when the LOVA happens during normal operation. Yet can not be observed when LOVA occurs during maintenance. The results are the basis to the further safety study of CFETR such as the re-suspension, migration and explosion of dust.
Proposed License Structure of Small Modular Reactor in China
Authors: Hua Zheng , Shuhong Wei
Published: October 2018

Small Modular Reactor (SMR) is getting more and more attention due to its safety and multi-purpose application. License structure is an important issue for SMR licensing. Modular design, construction and operation, shared or common structure, system and components (SSC) challenge existing large light water reactor license structure. Existing nuclear power plant license structure, characteristics of SMR and its effect on license structure, and research progress of U.S Nuclear Regulatory Commission (NRC) are analyzed, SMR license structure in China are proposed, which can be used as a reference for SMR R&D, design and regulation.
Penetration Behavior of Liquid Jet Falling Into a Shallow Pool
Authors: Kimura Fumihito , Yoshida Hiroyuki , Kaneko Akiko , Abe Yutaka
Published: October 2018

Mitigation measures against severe accidents (SAs) are important from the viewpoint of safety of nuclear reactors. In some scenarios of the SAs, the core materials melt and fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant is occurred. This process is called a fuel-coolant interaction (FCI).

The aim of the present study is to clarify the liquid jet behavior falling into a shallow pool. Our focus is on the atomization conditions of a liquid jet injected into the pool with insufficient depth.

In order to understand the jet behavior in a shallow pool, we performed observation of visualization with several methods and mapped observed flow regimes of jet against dimensionless numbers. As a results of observation, we succeeded visualization of internal flow.
Experiences on Radioactive Materials Safe Transport in CIRP
Authors: Jiangang Zhang , Guoqiang Li , Renze Wang , Hongchao Sun , Dajie Zhuang , Shutang Sun , Dongyuan Meng
Published: October 2018

The safety of radioactive material transport is regulated by graded approach in China. National requirements on container design, manufacture and shipment are different for the three categories material. Packages of different categories shall be experienced different test to demonstrate their ability to withstand transport normal transport conditions and accident conditions depending on their contents. China Institute for Radiation Protection (CIRP) established free drop test facilities, thermal test facilities, water immersion test facilities and relevant measurement means. CIRP has carried out tests of 18 radioactive material transport packages, and accumulate a lot of experiences about the limit unit analysis, test scheme, test process, monitoring and data treatment. Probabilistic safety assessment method and traffic accident statistics in China are carried out in the recent years by CIRP.
Electric Field Simulation of Ionization Chamber Used in Tritium Measurement in Tail Gas of Molten Salt Reactor
Authors: Qian Nan , Wang Guanghua , Yu Huang , Zeng Youshi , Bao Guangliang , Wu Shengwei , Liu Wei
Published: October 2018

There is a large amount of tritium in tail gas emitting from the molten salt reactor. The content of tritium content in tail gas must be measured and controlled according to the national standard. Nowadays, the ionization chamber and the liquid scintillation counter are mainly used for tritium monitoring. The proportional counter required pure sampling gas, and the liquid scintillation cannot measure the radioactivity online. Only the ionization chamber could measure the activities of off-gas from molten salt reactor. In order to reduce tritium contamination in the ionization chamber, a wire type high pressure ionization chamber was used in measuring tail gas from molten salt reactor. In order to improve the ion collection efficiency of the wire type electrode ionization chamber, it is necessary to optimize the number of the electrodes. In this study, the electric field of ionization chamber with different numbers of electrodes, such as 3, 6, 12, 18, 24, 30, 36, 42, 48, were simulated by COMSOL software. The equipotential line of electric field near the collector is round. The equipotential line of electric field near the high voltage wires present as undulate. From the results, the optimum number of wires was 36. The difference of electric field between two types chambers were less than 5%. After that, collection efficiency was simulated as well. Increasing the number of the high voltage electrode at top and bottom of effective area could not improve the electron collection efficiency. Making the ionization chamber shell as an insulator can effectively improve the electrons collection efficiency.
Development of Self-Reliant Thermal Hydraulic and Safety Analysis Software for NPP Design
Authors: Liu Yu , Deng Jian , Pan Junjie , Lu Zongjian
Published: October 2018

For the thermal hydraulic and safety analysis of nuclear reactor, a lot of theoretical models and engineering experience are mainly contained in the design software. The degree of self-reliance of software directly reflects the technological level and core competencies. According to the nuclear power plant (NPP) design requirements, self-reliant thermal hydraulic and safety analysis software, such as CORTH, TRANTH, PHYCA and etc., have been developed by Nuclear Power Institute of China (NPIC) with reasonable planning and scientific implementation. In this paper, the development process of the self-reliant software is reviewed, covering requirement analysis, model research, software design, coding, testing, verification and validation. And the main characteristics of self-reliant software were summarized. The successful development of thermal hydraulic and safety analysis software support the export of nuclear power units of China, and enhance the competitiveness.
Study of the Behaviors of an Open-Loop Heat-Removal System
Authors: Wang Xianmao , Shen Yonggang , Yang Jiang , Ouyang Yong , Rui Min , Qiu Suizheng
Published: October 2018

In the third generation of nuclear reactors, passive systems have been widely used such as passive core cooling system and passive containment cooling system, which usually relay on natural circulation induced by buoyancy force to remove heat. Most of these passive cooling systems are closed-loop natural circulations. In recent years, some open-loop heat-removal systems have also been put forward. Open-loop heat-removal systems have its own advantages such as its simplification and low costs. However, the thermal-hydraulic behaviors of open-loop heat-removal systems are still not totally clear and need further study.

In this study, a simplified open-loop passive containment cooling system is studied. A calculation model is built based on RELAP SCDAPSIM code. The thermal-hydraulic behaviors of the system are studied. By changing some key parameters of the system, the influences of these parameters on the system are evaluated.
Investigation on Fission Products Release Mitigated by In-Containment Relief Valve Under SGTR Accident
Authors: Kim Taeseok , Choi Wonjun , Jeon Joongoo , Kim Nam Kyung , Kim Sung Joong
Published: October 2018

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in Nuclear Power Plants (NPPs). However, in a bypass scenario of Steam Generator Tube Rupture (SGTR), radioactive nuclides are released to environment even if the containment is not ruptured. The radioactive nuclides are transported from primary to secondary systems through a broken steam generator tube during SGTR accident. Accordingly, the radioactive nuclides of the secondary system can be released to the environment through Main Steam Safety Valve (MSSV) or Atmospheric Dump Valve (ADV). Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise In-Containment Relief Valve (ICRV) from steam generator to the free space in the containment building of the Optimized Power Reactor 1000 MWe (OPR1000). This study focuses on the conceptual development of the mitigation strategy and MELCOR code was used for the numerical simulation. The MELCOR input model of OPR1000 consists of 58 control volumes and 161 flow paths. Safety features such as Pressurizer Safety Relief Valve (PSRV), Safety Injection Tanks (SITs), and MSSV were modeled in the MELCOR model. To initiate the SGTR scenario, a flow path between secondary and primary sides of Steam Generator (SG) was modeled with a flow area of 4.49 × 10−4 m2. The safety features were assumed that a few passive systems such as PSRV, MSSV, and SIT, were available. Under this condition, the ICRV connecting the SG and the free space in the containment such as dome and Reactor Drain Tank (RDT) were modeled. Specifications of the ICRV such as length, flow area, and valve opening condition were assumed to similar to those of the MSSV. Using these paths, three cases were considered; a base case, a case of steam release to the containment dome (CNMT case), and a case of release to the RDT (RDT case). Simulation results show that in the base case released radionuclides to the environment. In the other cases, the radioactive nuclides were not released to the environment although the containment pressure increased more than the base case, which is lack of the ICRV. As a result, the ICRV prevents the radionuclides release to the environment during SGTR accidents. Further studies are needed to incorporate practical valve inputs, reactor type, and safety features to gain more feasibility.
Experimental Research on the Flow Resistance and Heat Transfer Characteristics in Rod Bundle Channel
Authors: Zhu Zhiqiang , Tian Chunping , Yan Changqi , Wang Jianjun , Ren Tingting , Guo Zehua
Published: October 2018

Single-phase natural circulation experiments were conducted to study the flow resistance and heat transfer characteristics in a 3 × 3 rod bundle channel with the ratio of rod pitch and rod outer diameter (P/D) 1.38. The range of inlet subcooling degree is 30∼90K and the heating power is 1∼20kW. The rods are heated with constant heat flux.

According to the experimental results, the flow regime under natural circulation condition is divided and the transition Reynolds number is considered as 800. The flow transition is recognized by the slope change of friction factor curve since the flow transition in the rod bundle channel is not as obvious as that in round pipe. Simultaneously, the flow transition in the rod bundle is much earlier and the upper critical Reynolds number is much larger compared to regular channel like round pipe and rectangular channel. Two correlations for laminar and transition regime are fitted to calculate the friction factor. As for the grid spacer local resistance coefficient, there is slight change at Reynolds number 800 and similarly two correlations are fitted to calculate the local resistance coefficient. The Nusselt number tendency changes at around Reynolds number 4000 but keep unchanged at transition point, which means the flow transition has no obvious effect to the heat transfer. The heat transfer results are compared with different single-phase convective heat transfer correlations. D-B and Gnielinski correlations are not suitable for the heat transfer prediction in rod bundle channel and the relative deviation is more than 20%. Weisman, Presser and Markoczy correlations predict relatively well in high Reynolds number region, and Markoczy correlation is the best of them. In low Reynolds number region, most experimental results are larger than the correlations. D-B correlation based methods may be unsuitable for the heat transfer prediction in rod bundle channel and a new correlation needs to be proposed.