bluuebunny/crossref_metadata_2025_split 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 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.
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.
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.
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.
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.
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.
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.