bluuebunny/crossref_metadata_2025 huggingface dataset:
High Throughput Risk and Impact Screening of Chemicals in Consumer Products
Authors: Jolliet Olivier , Huang Lei , Hou Ping , Fantke Peter
Published: October 2020

The ubiquitous presence of more than 80,000 chemicals in thousands of consumer products used on a daily basis stresses the need for screening a broader set of chemicals than the traditional well‐studied suspect chemicals. This high‐throughput screening combines stochastic chemical‐product usage with mass balance‐based exposure models and toxicity data to prioritize risks associated with household products. We first characterize product usage using the stochastic SHEDS‐HT model and chemical content in common household products from the CPDat database, the chemical amounts applied daily varying over more than six orders of magnitude, from mg to kg. We then estimate multi‐pathways near‐ and far‐field exposures for 5,500 chemical‐product combinations, applying an extended USEtox model to calculate product intake fractions ranging from 0.001 to ∼1, and exposure doses varying over more than nine orders of magnitude. Combining exposure doses with chemical‐specific dose–responses and reference doses shows that risks can be substantial for multiple home maintenance products, such as paints or paint strippers, for some home‐applied pesticides, leave‐on personal care products, and cleaning products. Sixty percent of the chemical‐product combinations have hazard quotients exceeding 1, and 9% of the combinations have lifetime cancer risks exceeding 10−4. Population‐level impacts of household products ingredients can be substantial, representing 5 to 100 minutes of healthy life lost per day, with users’ exposures up to 103 minutes per day. To address this issue, present mass balance‐based models are already able to provide exposure estimates for both users and populations. This screening study shows large variations of up to 10 orders of magnitude in impact across both chemicals and product combinations, demonstrating that prioritization based on hazard only is not acceptable, since it would neglect orders of magnitude variations in both product usage and exposure that need to be quantified. To address this, the USEtox suite of mass balance‐based models is already able to provide exposure estimates for thousands of product‐chemical combinations for both users and populations. The present study calls for more scrutiny of most impacting chemical‐product combinations, fully ensuring from a regulatory perspective consumer product safety for high‐end users and using protective measures for users.
Power‐Sharing in Deeply Divided Societies: Consociationalism and Sectarian Authoritarianism
Authors: Dixon Paul
Published: October 2020

Consociationalists have traditionally embraced sectarian authoritarianism combined with a very limited form of democracy as the only democratic way of managing plural conflicts. New Generation consociationalists, by contrast, believe their theory is compatible with opposing sectarian authoritarianism in places like Iraq and the Lebanon. Traditional consociationalists have tended to claim that all power‐sharing is consociational, whereas revisionist and liberal consociationalists claim that consociationalism does not require power‐sharing and that the Iraqi Constitution of 2005 is therefore a consociational success. This paper argues that ‘classic’ consociationalism has constantly been revised to deflect criticism and capture apparently successful cases of conflict management. Consociationalists have deployed a ‘Wonderland’ definition which allows them to make such contradictory claims. Paradoxically, this definitional ambiguity and incoherence allows consociationalism to be ‘all things to all people’, resulting in its successful domination of the academic debate.
Anomalous Enhancement of Heat Transfer to H2O/CO2 Mixtures in Near-Critical Region
Authors: Zhang Hanlin , Wu Haomin , Li Sha , Liu Dong , Li Qiang
Published: October 2020

Heat transfer to supercritical H2O/CO2 mixtures (24 MPa, 310 to 430 °C, and CO2 mass fractions up to 18.5%), the working fluids of a novel power generation system with coal gasified in supercritical water, was experimentally investigated for typical working conditions of this system. For these conditions, i.e., high mass velocities (above 1200 kg m−2 s−1) and low heat flux (below 300 kW m−2), the convection heat transfer coefficients (HTCs) of supercritical pure fluids usually increase with temperature, peak near the pseudo-critical point, i.e., heat transfer enhancement, and then decrease for higher temperatures. Here, we experimentally demonstrated a new heat transfer enhancement phenomenon for supercritical H2O/CO2 mixtures. A high-temperature and high-pressure apparatus was setup to measure the convection HTCs of the supercritical H2O/CO2 mixtures. Experimental results show that surprisingly two distinct peaks of convection HTCs appear, with one corresponding temperature being the pseudo-critical point of the H2O/CO2 mixture, i.e., the thermophysical property variation induced mechanism, and the other one being the critical miscible point of the mixture, i.e., the dissolution-induced mechanism. These results pave the way to efficient heat transfer devices that use supercritical mixtures as heat transfer fluids.
Nuclear Pressure Vessel Manufacture Using the Hot Isostatic Pressing (HIP) Process
Authors: Sulley John , Wallace Phil , Warner Ted , Jones Gary
Published: October 2020

Hot Isostatic Pressing (HIPing) has been used by Rolls-Royce to successfully manufacture nuclear plant components such as valves, piping, and pump casings; the majority of these components being manufactured from stainless steels, typically 316L. There are also considered to be potentially significant benefits to be gained by manufacturing large nuclear plant pressure vessels via the HIP process, such vessels commonly being manufactured from Low Alloy Steel (LAS) materials such as ASME SA-508. The benefits would include cost and lead-time reductions, which are particularly pertinent in relation to the competiveness of the power generation market and future nuclear power plant construction. Such vessels are a major cost and are critical path items of the primary plant. Also, material quality improvements and improved inspectability are possible via the HIP process. Welding vessel sections together using Thick-Section Electron Beam Welding (TSEBW) shows significant promise in reducing welding time and the provision of high quality welds, further reducing vessel cost and lead-time. There is also the potential with the use of TSEBW, to reduce weld inspection requirements with the weld being effectively the same as the parent material, i.e. no weld filler material is used. This paper presents an overview of the pioneering work conducted and planned by Rolls-Royce to develop a method of manufacture to combine HIPing and TSEBW to produce nuclear plant pressure vessels. Staged development is covered, starting with small billet manufacture for the purposes of material testing and examination, followed by vessel demonstrators for the purposes of proving the method of manufacture and to provide justification data, e.g. examination, pressure and thermal cyclic test data.

In order to provide a balanced perspective, the paper also identifies the key challenges — risks, and capability development requirements necessary to deliver this method of manufacture.
Investigation of VVER-440 Confinement Response During DBA After Leaktightness Improvement
Authors: Hovhannisyan Hovhannes , Malakyan Tsolak , Amirjanyan Armen
Published: October 2020

Based on Armenian Nuclear Power Plant (VVER-440/270) example the response of confinement to leaktightness improvement in case of design basis accident was investigated. The main interest of the investigation is to determine the influence of confinement leaktightness improvement on the reduction of available NPSH of active safety systems pumps (ECCS and spray systems). Results of analyses showed that confinement leaktightness improvement will lead to strong underpressure condition and endanger ECCS pumps operation due to cavitation. Hence, in parallel with confinement leaktightness improvement confinement vacuumization prevention means must be considered.
Emergency Action Level Study in Reprocessing Facility
Authors: Renze Wang , Jiangang Zhang , Yapeng Yang , Zongyang Feng , Linsheng Jia , Ning Wang , Yining Liu , Boning Liang
Published: October 2020

The safety characteristics and potential hazards of reprocessing facilities are different from those of nuclear power plants (NPPs). Emergency action level (EAL) development is an important aspect of emergency preparedness, and EAL is an important basis for emergency response of reprocessing facilities. EAL quantitative research can enhance its operability.

At present, the domestic and foreign literature, generally only gave the principle method for EAL development. There are no operational guidance documents for specific EAL quantification in reprocessing facilities. According to the features of the functions and configurations of the reprocessing facilities, two additional categories EAL, E category for the spent fuel pool accident and W category for the high level liquid waste tank accident, have been added. Meanwhile, four categories-S, F, A, H are retained with similar implication as NPPs.

On the basis of existing principles and practical experience, specially the reference from DOE G 151.1-1A, EAL quantification was developed according to the characteristics and symptoms related to the safety of reprocessing facilities.

EAL quantification for several accidents was developed, and it was with good maneuverability. The study for EAL quantification in reprocessing facilities shows that, quantitative, indicative, practical EALs can make emergency response more accurate and efficient, and ongoing research is strongly requisite.
Modeling the Molten Salt Reactor Experiment With the GOTHIC Code
Authors: Harvill Rodney , Lane Jeff , Link John , Gates Anita , Kindred Tom
Published: October 2020

GOTHIC 8.3(QA) includes capabilities for modeling advanced, non-light water cooled reactors. Important capabilities introduced in GOTHIC 8.3(QA) include fluid property tables for various molten salts, an enhancement to the tracer tracking module to allow radioactive decay energy to be released locally in the carrier fluid and other improvements to the neutron kinetics module. With these new capabilities in place, GOTHIC is used to benchmark steady-state and transient conditions in the Molten Salt Reactor Experiment (MSRE), which operated at Oak Ridge National Laboratory from 1965 to 1969. In this experimental reactor, UF4 fuel was dissolved in molten fluoride salt, and criticality could be achieved only in the graphite moderated core. An air-cooled radiator transferred fission and decay heat to the environment.

The design thermal output of the MSRE was 10 MWt, but the radiator design limited the output to 8 MWt. The original design parameters neglected the impact of decay heat on system temperatures. GOTHIC is used to benchmark system operating parameters at both the 10 MWt design condition and the 8 MWt operating condition, both with and without decay heat. The cases that include decay heat apply 7% of the nominal thermal output using the eleven decay heat precursors from ASB 9-2 as tracers. The results of the benchmark exhibit good agreement with design and operating data and demonstrate heat-up due to decay heat in the fuel salt outside the core.

In the MSRE, delayed neutron precursors are not confined to the core because the fuel and fission products flow through the system. As a result, there are different values for (effective) delayed neutron fraction with and without flow, and the decay of delayed neutron precursors outside the core under full-flow conditions reduces reactivity by 0.212 % δk/k. Zero power physics testing included fuel salt pump start-up and coast-down transients with a control rod automatically moving to maintain criticality. The control rod motion calculated by GOTHIC is a reasonable match to measured data from these transients.

Low power testing included a natural convection transient with no control rod motion such that reactor power was responding to heat load demand from the radiator. The reactor power and fuel salt and coolant salt temperatures calculated by GOTHIC exhibit good agreement with measured data.
Identify the Broken Loop and Break Sizes of Recirculation Line LOCA for a BWR6/MarkIII Plant
Authors: Hwang Sheng-Dih , Wang Te-Chuan
Published: October 2020

In this study, the MAAP5 code was used as the tool to simulate a reactor operated at steady-state and occurred an LOCA with different break sizes at the recirculation loops. A BWR6/MarkIII nuclear power plant (NPP) was selected as the sample deck. By detecting the temperature difference between the loops, and then the break size could be determined including which loop and how is the break size. Moreover, this study had also analyzed the effect of break sizes affecting the characteristic of an NPP. These results were consistent with the previous experiments and codes. This study also suggested to the operators in the main control room (MCR), when they observed some syndrome of the acquired data and they could preliminary estimate which loop broke and how was the size of the LOCA. Those could help the operators to make a strategy to avoid the probability of core melted.
The Experimental Research on the Mechanical Characteristics of the Newly Designed Fuel Rod Supporting Structure of Spacer Grid
Authors: Ren Quan-yao , Pu Zeng-ping , Zheng Mei-yin , Chen Jie , Qin Mian , Han Yuan-Ji , Peng Yuan , Chen Ping , Zhao Rui-rui , Li Jian-wei
Published: October 2020

The fuel rod supporting structure of spacer grid is the key to ensure the horizontal and axial position of the fuel rod. On the purpose of improving the neutron economy and simplifying the manufacturing process of the spacer grid, the innovative curved dimple and arched spring have been designed, which could be directly punched from the strip of spacer grid. The mechanical experiments have been performed to acquire the deformation-load curves of dimples and springs on single strap and in-cell supporting structure, which could provide the load and residual deformation at 100% and 120% designed nominal deformation. It has been demonstrated that the designed in-grid cell has a relative stable load at the 100% nominal designed deformation and the test method of the single stripe is representative for the similar load-deflection curve between the spring and the in-grid cell.
Research on Fault Diagnosis of Reactor Coolant Accident in Nuclear Power Plant Based on Radial Basis Function and Fuzzy Neural Network
Authors: Sun Pengpeng , Liu Yong , Wu Guohua , Duan Zhiyong
Published: October 2020

Nuclear power plants (NPPs) are widely used in the world. After three nuclear accidents, people propose higher of the safety and reliability on NPPs. Reactor coolant system (RCS) in the NPP directly affects whether the heat can be exported and radioactivity can be inclusive. It plays an important role of the NPPs safety. So, it is great significance of fault diagnosis for RCS in NPP.

Although many scholar had carried out research on fault diagnosis of NPPs, different networks may lead to different results in a system. Therefore, this paper chooses a system and uses different neural networks (NN) for comparative analysis which can provide advice for follow-up research. In the paper, RCS has been analyzed and typical fault have been analyzed through PCTRAN simulator. On this basis, two kinds of NN combined with fuzzy systems: radial basis function (RBF) and back propagation (BP) are used for fault diagnosis and comparative analysis. Loss of coolant accident, single pump failure, loss of feed water are set for simulation experiment. Simulation experiment shows that BP network’s hidden layer nodes is less than RBF-NN, but iteration speed of BP network is faster; accuracy of fault diagnosis based on BP-NN is higher than RBF-NN; fuzzy-NN for fault diagnosis is faster than NN.
Implementation and Validation of an Aerosol Collection Model by a Spray in a CFD Code: Application to the Scavenging of Aerosols Released During Laser Cutting Operations of Fuel Debris for the Dismantling of the Damaged Reactors of Fukushima Dai-ichi
Authors: Gelain Thomas , Porcheron Emmanuel , Leblois Yohan , Doyen Ioana , Chagnot Christophe , Journeau Christophe , Roulet Damien
Published: October 2020

The general context of this article is related to the dismantling of the damaged reactors of Fukushima Dai-ichi and, more specifically, to the implementation of the laser cutting technique for the fuel debris retrieval. IRSN is involved in a project led by ONET Technologies and in partnership with CEA, to bring relevant elements in order to analyze the risks induced by the dispersion of aerosols released by the dismantling operations.

During the laser cutting operations in air or underwater conditions, particles will be produced, involving a potential risk of dispersion into the environment. Hence, in order to prevent this situation, their collection is one of the safety key issues in the in-situ dismantling actions. For that, IRSN performed CFD simulations of aerosol scavenging by a spray to evaluate the collection efficiency by this technique.

These simulations, conducted with the ANSYS CFX code, use an Eulerian method for the continuous phase, and a Lagrangian method for the spray for which a collection model detailed by Plumecocq [1] or Marchand [2] was implemented. Aerosols are modelled by a DQMOM population balance implemented by Gelain et al. [3] (already used for recent simulations in the same context), and enriched with a deposition model developed by Nerisson et al. [4].

At first, CFD simulations were performed with the geometry of the IRSN TOSQAN facility [5], comparatively to experimental results presented in a previous paper [6]. This step enables the validation of the collection model implementation and to study the sensitivity to the aerosol size.

Then, CFD simulations were conducted with the geometry of the pedestal of Fukushima Dai-ichi reactors, to be more representative of a realistic case. For this configuration, sensitivity studies are described, highlighting both the influence of a multispray and of thermal-hydraulic conditions (temperature) on aerosol scavenging efficiency.
Study on the Spacer Grid Dynamic Crush Strength According to Cell Sizes
Authors: Ryu Jooyoung , Woo Hangil , Nam Chaeyoung , Yoo Youngik , Park Namgyu , Lee Seungjae
Published: October 2020

A spacer grid is one of the primary components of the PWR nuclear fuel. Spacer grid maintains proper pitches between the fuel rods and enables the fuel rod to cool down by providing coolant flow path. However, when the nuclear fuel is subjected to an unexpected excessive load during shipping, handling, manufacturing and operating, it could lead to fuel failure such as spacer grid buckling and cladding tube deformation. The most limiting load acting on the spacer grid is the lateral impact load during seismic/loss-of-coolant accidents. Dynamic crush strength of the spacer grid greatly contributes to the nuclear fuel integrity throughout the fuel lifetime [1]. This buckling strength tends to become weak in end of life (EOL) condition. KEPCO NF (KNF) carried out dynamic crush tests of the spacer grid and analyzed its characteristics. Spacer grids were prepared with three groups that have different cell sizes according to beginning of life (BOL), EOL and enlarged EOL simulated conditions. In addition, two kinds of dynamic crush tests were performed. One is pendulum impact test that drops a hammer to the grid in a short time. And the other is hydraulic long-pulse test that pushes impact plate to the grid in longer time. These tests and analysis results were compared in each group and discussed to explore key factors for improving crush strength of the spacer grid. In this paper, the spacer grid manufactured by additive manufacturing (AM) technology [2] is also introduced to verify the buckling performance. AM is a method to make designed shape with metal powder and built-up technology that is different from conventional manufacturing. Through the study, it could be a good alterative solution that the new manufacturing method might be helpful to improve dynamic impact characteristics.
Study of Laser 3D Scanning Model Reconstruction for Nuclear Facilities Decommissioning
Authors: Yongling Zhang , Wan Zhao , Shuai Wang
Published: October 2020

The laser 3D scanning model reconstruction is an effective technical means for the reconstruction of 3D virtual scenes of old nuclear facilities. However, the existing technology have the following problems: First, the scanning equipment needs to be installed on the triangular fixed bracket, which requires personnel to be installed on site, thereby bringing radiation safety risks to personnel; second, it requires the person to reconstruct and splice the model one face by one face, which is inefficient. Aiming to deal with these problems, this paper proposes a mobile laser 3D scanning method and a rapid model reconstruction technology. These methods include: (1) In order to realize fast scanning of large-scale scenes, and at the same time ensure accurate scanning of local important objects, a laser three-dimensional composite scanning method is proposed. (2) Design and development of a remote control automatic walking and fixing device, which can transport the laser 3D scanner to the position of measurement which is needed. The device has a certain ability to overcome obstacles, and can be reliably fixed to ensure the stability and scanning accuracy of the laser scanner at high speed working station. You can remotely control the laser scanner to start or stop working. The difficulty of this technology is how to make sure the device has the flexibility in motion and the stability on working; (3) An automatic model reconstruction and feature identification method based on a large number of point cloud data is proposed, and the corresponding software is developed. This method can realize the automatic recognition of the standard and common models, and the automatic model reconstruction of the point cloud data in the scene. To achieve this goal, the neural network algorithm is used. These technologies can effectively reduce the radiation safety risk of personnel during laser scanning in high-radioactive places, reduce the intensity of personnel operations, and improve work efficiency.
Point Defect Effects on Tensile Strength of BCC-Fe Studied by Molecular Dynamics
Authors: Lin Pandong , Nie Junfeng , Liu Meidan
Published: October 2020

BCC-Fe is the critical and major component of the reactor pressure vessel (RPV) steel. With long-tern neutron irradiation, many point defects can be obtained in RPV steel. In this paper, the points defects (interstitial, vacancy and Frenkel pair) effects on the tensile strength of Fe are studied by molecular dynamics simulations at 300K. The uni-axial tensile load is along [001] direction of the Fe samples loading in constant strain rate. The Fe atoms are added or removed randomly to generate point defects. For point defects, three types of point defects can decrease the tensile strength containing yield stress and strain of Fe samples. In addition, the tensile strength decreases with the increase of point defect concentration. With the same defect concentration, interstitials decrease the yield stress the most seriously compared with the vacancies and Frenkel pairs. Apart from that, the morphology and evolution of the microstructure of Fe with point defects are also investigated under tension. Compared with the perfect crystal, the generation of dislocation decreases the tensile strength dramatically. For sample with interstitials, interstitial clusters form and evolve in dislocations loops finally. For sample with vacancis, vacancy may aggregate together and vacancy clusters form as a result, which is seen as precursors of dislocation loop. Notably, the results are meaningful to understand the effects of point defects on tensile strength of BCC-Fe.
Effects of Rolling on Characteristics of System Under Forced Circulation and Natural Circulation
Authors: Cai Rong , Yue Nina , Fang Hongyu , Chen Baowen , Liu Lili , Ma Zehua
Published: October 2020

The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.
Nuclear Facility Safety at the United States Department of Energy
Authors: Frias Patrick , Muñoz José R. O. , Restrepo Louis , Tingey James L. , Louie David L. Y.
Published: October 2020

Nuclear facility safety is crucial to preventing and/or reducing high consequence-low probability accidents and, thus reducing the potential risks posed by United States Department of Energy (DOE) and National Nuclear Security Administration (NNSA) operations at their facilities/activities. DOE/NNSA has the responsibility of developing, issuing, maintaining, and enforcing nuclear safety Directives while fostering a culture that promotes nuclear safety research and development. Lessons learned from past accidents, near misses, and experiments/analyses are also important resources for improving operational nuclear safety in the safety community.

This paper first identifies and describes the current Directives in place, including safety review and regulatory process, and safety programs that support implementation of the Directives. This paper also describes a contractor’s approach to identifying and implementing safety using these Directives and lessons-learned in multiple discipline areas of nuclear safety.
Flow Pattern Transition Criteria for Upward Two-Phase Flow in Annulus
Authors: Wen He , Chenru Zhao , Hanliang Bo
Published: October 2020

Vertical upward two-phase flows in annulus are of great importance in many industrial fields due to the closely relationship between the flow patterns and the heat transfer characteristics. Common flow patterns in annulus are bubbly (B), slug (S), churn (C) and annular (A) flow, most of which are quite similar to those in tubes. However, due to the elliptic nose and asymmetric shape of the Taylor bubble in annulus, the slug to churn flow transition could be influenced by the channel geometry which was usually ignored in most of the previous researches. The flow pattern transition criteria for tubes are thus not applicable for annulus, especially for slug to churn flow transition, which should be separately studied. Therefore, in this paper, the basic characteristics of the flow pattern in annulus and their transition mechanism are analyzed. In addition, a set of semi-empirical transition criteria with higher accuracy are assessed and selected for annulus based on theoretical analysis and comparisons with experimental data.
Research on Fault Diagnosis Based on Hierarchical Signed Directed Graph for Nuclear Power Plants
Authors: Guohua Wu , Jiaxin Wang , Diping Yuan , Yiqing Xiao
Published: October 2020

When nuclear power plants (NPPs) are in failure state, it may release radioactive substance into the environment. Thus, the safety of NPPs is put forward a high standard. Fault detection and diagnosis (FDD) are significant for NPPs to help operator timely know the state of system and provide the online guidance. Fault diagnosis can improve the safety of nuclear power plants, but current fault diagnosis methods pay too much attention to accuracy of diagnostic results. As a complex industrial system, how to explain the causes of faults in NPPs becomes more important. Although there are many studies on the knowledge graph, there is no detailed description on the failure process (consider timing). This paper proposed a three-layer structure for FDD in NPPs. Each layer represents the stage of the accident, it can give the operator a clear cognitive process to faults. The three-layer structure includes “smooth layer”, “threshold layer” and “fault layer”. The three layers indicate the reason of faults, the response of the parameters at each stage, and clearly showing the accident process. The smooth layer uses the stability analysis to analyze whether the current NPP is operating abnormally; the threshold layer uses the thresholds of the NPP to monitor which parameters have exceeded the upper limit or the lower limit; the fault layer reflects what is happening in the current operation and accidents are explained using signed directed graph. This paper takes the loss of coolant accident as an example, three-layer structure is analyzed, which shows the feasibility of the method. The case shows that the proposed method is superior to the conventional SDG method, can diagnose the faults, and give the reason of diagnosis results.
Methodology Development for Transient Flow Distribution Analysis in High Temperature Gas-Cooled Reactor
Authors: Aoki Takeshi , Sato Hiroyuki , Ohashi Hirofumi
Published: October 2020

In the thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR), unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed to minimizing the unintended flows. The flow distribution considering unintended flows in the reactor has been evaluated for steady and conservative condition. On the other hand, the transient thermal hydraulic analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR.

The present study aims to improve the thermal hydraulic system analysis code developed by Japan Atomic Energy Agency based on the RELAP5/MOD3 code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents for its design improvement utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated to evaluate the detailed flowrate distribution considering the unintended flows in the core and the molecular diffusion that is important to analyze beginning air ingress behavior in an air ingress accident triggered by a rupture of a primary coolant piping in HTGR. It is concluded that a prospect has confirmed to apply the improved thermal hydraulic system analysis code for transient flow distribution analysis for prismatic-type HTGRs.
Experimental Study of Pure Steam and Steam-Air Mixture Condensation on a Vertical Chrome-Plated Tube
Authors: Niu Zesheng , Fan Guangming , Cheng Jie , Li Wei
Published: October 2020

To investigate the heat transfer characteristics of the chrome-plated tube and hope it could be used on the internal heat exchanger of passive containment cooling system (PCS), an experimental investigation has been conducted. In this experiment, a series of steam condensation experiments are performed under pure steam and steam-air mixed conditions over chrome-plated tube for a variety of chromium coating thickness (1μm and 10μm), total pressure, air mass fraction and wall subcooling. Condensation heat transfer coefficient was obtained for the total pressure ranging from 0.2 MPa to 0.4 MPa, air mass fraction ranging from 0.10 to 0.71, and wall subcooling from 10°C to 70°C. Moreover, the designed visualization experimental device makes the experimental phenomenon can be directly observed through the observation window. Under the pure steam condition, the results show that droplet condensation and filmwise condensation is co-existed on both two kinds of chrome-plated tubes, the chrome coating thickness of 10μm tube shows better heat transfer ability. Under the steam-air mixed condition, the condensation heat transfer coefficient of both two kinds of tubes increases with total pressure, and decrease with the air mass fraction and wall subcooling, while the influence of chrome coating thickness on heat transfer is no longer noticeable. The results also indicate that the thickness of the chromium coating will affect the surface microstructure of the chrome-plated tube and then affect the heat transfer ability of the chrome-plated tube.
The Fire Event Analysis for Fire Frequency Estimation on Japanese Nuclear Power Plant
Authors: Nagata Yasunori , Uchida Tsuyoshi , Shirai Koji
Published: October 2020

To promote the practical fire Probability Risk Assessment (PRA) among Japanese industries based on the domestic fire experience, it is important to estimate generic fire ignition frequency distributions for Japanese nuclear power plants (NPPs), which are applicable to the fire PRA during full power operation.

Firstly, the Central Research Institute of Electric Power Industry (CRIEPI) proposed the estimation process of the distributions, which consists of the four steps including sub steps as follows:

Step 1: Fire event data collection including development of criteria

Step 2: Fire event data analyses including development of criteria:

Step 2.1: Screening of fire events to identify events to be classified

Step 2.2: Classification of fire events from the viewpoint of fire severity of each event.

Step 2.3: Categorization of fire events from the viewpoint of fire ignition source types

Step 3: Estimation of generic fire ignition frequency distributions

Step 4: Comparison of the results of Step 3 with fire ignition frequency estimated in other countries

In this paper, CRIEPI introduced the detail of Step 1 and 2 above and discussed the study results for them, which are essential to estimate generic fire ignition frequency distributions properly. Through Step 1, CRIEPI established a preliminary fire event database based on fire records at Japanese NPPs provided by domestic utilities. In step 2.1, CRIEPI identified the 180 fire events for future steps. Finally, CRIEPI classified the fire events and defined the fire ignition source categories in Step 2.2 and 2.3. As a result, CRIEPI found the major three risk contributors among fire ignition source categories: transient related fires, electrical cabinet fires and pump fires.
Thermo-Mechanical Analysis of Instrumentation Guide Tube Failure During a Severe Accident in a Nordic Boiling Water Reactor
Authors: Yue Ying , Villanueva Walter , Wang Hongdi , Wang Dingqu
Published: October 2020

Vessel penetrations are important features of both pressurized water reactors and boiling water reactors. The thermal and structural behaviour of instrumentation guide tubes (IGTs) and control rod guide tubes (CRGTs) during a severe accident is vital in the assessment of the structure integrity of the reactor pressure vessel. Penetrations may fail due to welding failure, nozzle rupture, melt-through, etc. It is thus important to assess the failure mechanisms of penetrations with sufficient details. The objective of this paper is to assess the timing and failure modes of IGTs at the lower head during a severe accident in a Nordic boiling water reactor. In this study, a three-dimensional local finite element model was established using Ansys Mechanical that includes the vessel wall, the nozzle, and the weld joint. The thermo-mechanical loads of the finite element model were based on MELCOR results of a station blackout accident (SBO) combined with a large-break loss-of-coolant accident (LBLOCA) including an external vessel cooling by water as a severe accident management strategy. Given the temperature, creep strain, elastic strain, plastic strain, stress and displacement from the ANSYS simulations, the results showed the timing and failure modes of IGTs. Failure of the IGT penetration by nozzle creep is found to be the dominant failure mode of the vessel. However, it was also found that the IGT is clamped by the flow limiter before the nozzle creep, which means that IGT ejection is unlikely.
Sensitivity and Uncertainty Information Incorporated Loss of Flow Accident Analyses for Research Reactors
Authors: Liu Tao , Wu Zeyun
Published: October 2020

This paper outlines a system level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The protected loss of flow (LOF) accident was selected as an exemplified design basis accident to demonstrate the analysis procedure. The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor was adopted as a research reactor model in the study. Two system level dynamics codes, RELAP5-3D and PARET, were employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the system code models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios. The canonical transient predictions on the LOF accident yielded from the two system codes mentioned above have demonstrated some noticeable yet acceptable discrepancies. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the RELAP5-3D model and the data analytic engines provided by the RAVEN framework developed by INL. The sensitivity information reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions delivered by the framework.
Discussion of Water Quality Factors Affecting 137Cs Adsorption and the Relationship Between 137Cs-Kd and Flow Rate in Freshwater
Authors: Wang Jiaxin , Zhang Liuguo , Qu Jingyuan , Wu Guohua
Published: October 2020

The discharge water of Inland Nuclear Power Plant (INPP) is freshwater. Freshwater has a lower environment capacity and more complex environmental impacts comparing with the salt water (oceans), which makes the public more sensitive and stricter. 137Cs, with a long half-life of 30.17 years, is likely to exist as particle form in freshwater, which increases the difficulty to be removed and also becomes a long-term potential radiation hazard. The factors affecting the behavior of 137Cs in freshwater could be divided into water quality factors and hydraulic factors. For water quality factors, the paper summarizes factors that have significantly influence on the adsorption of 137Cs: size and concentration of suspended particles, competing cations, pH, concentration of organic matters and another two environmental factors, time and extreme meteorological events such as floods, heavy rain. We describe the qualitative relationship that tells how the factors take effect on adsorption process. In terms of hydraulic factors, this paper focuses on a representative factor, flow rate, and takes the original measured distribution factor (137Cs-Kd) and flow rate data of 9 freshwaters from Asia and Europe for analysis. The study finds that there are two segmented index relationships between 137Cs-Kd and flow rate above the threshold of 1 m3/s, the experiences of other five rivers that had not participated in data fitting verify the accuracy and generic character of the index relationships. The qualitative summary of water quality factors and the quantitative conclusions between 137Cs-Kd and flow rate can provide advice for evaluating the behavior of 137Cs in freshwater, and then help in the site selection process for INPP.
Experimental Investigation on Temperature Characteristics of Direct Contact Condensation in a Natural Circulation System
Authors: Sun Jianchuang , Ding Ming , Mi Zhengpeng , Zhang Zhuohua
Published: October 2020

Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 °C.
Study on Irradiation Damage of RPV Steels Based on Coupling Cluster Dynamics and Crystal Plasticity Finite Element Method
Authors: Wang Xiaotong , Luo Ying , Dong Yuanyuan , Yao Weihua
Published: October 2020

Irradiation hardening is one of the most important aging effects of reactor pressure vessel (RPV) steel in long-term service. A number of studies have indicated that irradiation hardening is mainly caused by irradiation induced defects, such as dislocation loops and precipitates. In this paper, we have simulated the irradiation damage of low-copper reactor pressure vessel (RPV) steel. The generation of Mn-Ni-Si precipitates is simulated by the cluster dynamics model based on rate theory. On this basis, the crystal plasticity finite element method based on modified crystal plasticity model is used to simulate the effect of Mn-Ni-Si precipitates on the mechanical properties of RPV steels. The simulated results has been compared with the experimental results from the literature. By coupling the cluster dynamics and the crystal plastic finite element method, we suggest a multi-scale simulation method to simulate and predict irradiation damage of RPV steel.
Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR
Authors: Jiang Baihui , Zhou Zhiwei , Xia Zhaoyang , Sun Qian
Published: October 2020

Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors.

RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators.

In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.
Impact Analysis of NPP H4 Connections Design Improvement on Emergency Operation
Authors: Yuqi Wang , Ke Yi
Published: October 2020

After the loss of coolant accident (LOCA), the safety injection system injects water into the reactor coolant system (RCS), and the residual heat rejects from the break. The containment spray system is operating in recirculating cooling mode to ensure that the containment is cooldown. This state must be maintained for several months. After the accident, in order to respond the design extension conditions (DEC) of failure of two containment spray pumps or two low pressure safety injection pumps, the design of the original H4 connections was improved, and the H4 procedure (loss of containment spray pumps or low pressure safety injection pumps) was developed. H4 procedure demands to put into service 2 permanent (one for each train) interconnections of containment spray system and safety injection system, called “H4 connections”. Through the design improvement of the H4 connections, the mutual backup function of safety injection and containment spray can be realized implemented. The manual valves of the H4 connections in the original design were changed to electric valves, which ensured the accessibility of operator and avoided the radiation of high radioactivity level to operator after the accident. In addition, the improved H4 connections enable mutual backup of safety injection and containment spray in the early stage after the accident to be implemented, which fully improves the ability to respond to accidents and safety design level of the nuclear power plant (NPP). This also makes it possible to intervene early after the accident.