bluuebunny/crossref_metadata_2025 huggingface dataset:
Setup of the Supercritical CO2 Test-Facility “Scarlett” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System
Authors: Flaig Wolfgang , Mertz Rainer , Starflinger Jörg
Published: October 2017

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. This can lead to the development of more compact and more efficient components, e.g. heat exchanger and compressors.

A specific field of interest is a new decay heat removal system for nuclear power plants which is based on a turbine-compressor-system with supercritical CO2 as the working fluid. In case of a station blackout this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. This scenario has already been investigated by means of the thermo-hydraulic code ATHLET, numerically demonstrating the operation of this system for more than 72 h. The practical demonstration is carried out within the Project “sCO2-HeRo”, funded by the European Commission, in which a small scale demonstration unit of the turbo compressor shall be installed at the PWR glass model at GfS, Essen, Germany. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger needs to be as compact and efficient as possible. Therefore, a diffusion welded plate heat exchanger (DWHE) was developed and manufactured at IKE. It has been designed with rectangular mini-channels (0.5–3 mm hydraulic diameter) to ensure high compactness and high heat transfer coefficients. Due to uncertainties the DWHE has to be tested in regard to the actual possible transferrable heat power and to the pressure loss. According to this demand a multipurpose facility has been built at IKE for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to the test section. The test section itself can be exchanged by other ones for various investigations. After the test section, the CO2 pressure is reduced and the liquid is stored in storage tanks, from where it is evaporated and compressed again. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa and temperatures up to 150 °C. The first subject of interest will be the study of the thermal behavior of a DWHE using supercritical CO2 as a working fluid close to its critical point. Experiments concerning pressure loss and heat transfer will be carried out as a start for fundamental investigation of heat transfer in mini-channels. This paper contains a detailed description of the test facility, of the first test section and first results regarding heat transfer power and pressure loss.
Requirement and Discussion of Design Extension Conditions
Authors: Jinquan Yan , Song Chen , Lin Tian , Minglu Wang
Published: October 2017

Nuclear safety especially severe accidents risks are of great concerns of nuclear power plant. Design consideration of severe accident prevention and mitigation is generally required by various nuclear safety authorities worldwide. However, those requirements related to severe accidents consideration are somewhat different from country to country. Recently, the International Atomic Energy Agency (IAEA) updated and published a safety code on Specific Safety Requirement of Nuclear Power Plant Safety: Design (SSR-2/1). Meanwhile, the Chinese National Nuclear Safety Administration (NNSA) also revised and updated the safety code on Requirement of Nuclear Power Plant Safety in Design (HAF102). In these two codes, both IAEA and NNSA established some new requirements, among which two are of great concern. One is Design Extension Conditions (DEC) for consideration of those conditions traditionally called Beyond Design Basis Accidents (BDBA) in design of nuclear power plant, another is requirement of practically eliminating large release of radionuclide. These two new requirements are internally related, somewhat different and more restrict from those related to severe accident requirements set forth by Nuclear Regulatory Committee of United States (USNRC). Up to date, there are no specific guidelines about engineering implementation of those new safety codes. This paper present an overview of those requirements from IAEA, WENRA, NRC and China NNSA, followed by discussion of engineering approach for the implementation of the DEC requirement set forth by safety authorities.
Physical Design on PMNSR With LEU Core
Authors: Xiaobo Wu , Dan Peng , Jingyan Hong , Jin Lu , Qian Hao
Published: October 2017

Prototype Miniature Neutron source Reactor (PMNSR) is a pool-tank type research reactor,applying high enrichment Uranium as fuel, light water as moderator and coolant, beryllium as reflector. Recently, in order to prevent nuclear proliferation, PMNSR carried out low enrichment uranium (LEU) core conversion, and the enrichment of U-235 decreased from 90% to under 20%. Research on PMNSR with LEU core mainly includes theory design, zero power experiment, core replacement. The physical design of PMNSR with LEU is the main part of theory design, which plays a great role in LEU conversion. At the first stage of LEU conversion, it performs preliminary physical calculation and analysis concerning U-235 fuel enrichment, and the number of critical fuel elements, the reactivity worth of control rod, the reactivity worth of top beryllium reflector, the neutron flux of inter-irradiation tube are calculated, which provides important data for the fuel elements design, fabrication, zero power test safety analysis and experiment for LEU conversion. In the second phase, it conducts the result verification on zero power test and preliminary physical design and a preliminary error analysis resulted from it thereof. More over, it modifies input file of LEU conversion, optimizes core element loading deployment, the reactivity worth of central control rod, the neutron flux rate of inner radiation site, offering statistics for the replacement and start-up experiments. In the last period, grounded on the counting abnormal analysis in loading, it explains the reasons with calculation results, completing PMNSR LEU conversion. PMNSR physical design takes the leading position in LEU conversion. It supplies reference data to ensure completion of PMNSR conversion and lays a theoretical foundation for Ghana and Nigeria MNSR LEU core conversion.
Study on Semi-Analytical Nodal Method for Solving SP3 Equation
Authors: Qingquan Pan , Haoliang Lu , Dongsheng Li , Kan Wang
Published: October 2017

Solving the SP3 equation is the key technology of the Next Generation Reactor Physics Calculation, and has been widely concerned. The semi-analytical nodal method (SANM) based on transverse-integrated neutron diffusion equation has the advantages of high accuracy and convenience for multi-group calculation. The 0th-order flux and the 2nd-order flux being Expanded with the existing 4th-order SANM polynomials and being solved respectively, the 4th-order algebraic accuracy flux distribution is also obtained, however, this solving process is not the semi-analytical nodal method since the polynomial expansion process does not take the special modality of SP3 equation and it’s analytical solution into consideration. There are two modality SP3 equation, so there are two SANM expansion forms. A code is developed to solve the SP3 equation under the two different forms. After the calculation of the same benchmark, the difference between the two forms of SP3 equation is found. According to the results, and in view of the special modality of the SP3 equation, advices for a strict semi-analytical method for solving SP3 equation are discussed.
Biaxial-EDC Test Attempts With Pre-Cracked Zircaloy-4 Cladding Tubes
Authors: Li Feng , Mihara Takeshi , Udagawa Yutaka , Amaya Masaki
Published: October 2017

When the pellet-cladding mechanical interaction (PCMI) occurs in a reactivity-initiated accident (RIA), the states of stress and strain in the fuel cladding varies in a range depending on the friction and degree of bonding between cladding and pellet. Japan Atomic Energy Agency has developed the improved Expansion-due-to-compression (EDC) test apparatus to investigate the PCMI failure criterion of high-burnup fuel under such conditions. In this study, the failure behavior of cladding tube was investigated by using the improved EDC test apparatus. Cold-worked, stress-relieved and recrystallized Zircaloy-4 tubes with a pre-crack were used as test specimens: this pre-crack simulated the crack which is considered to form in the hydride rim of high-burnup fuel cladding at the beginning of PCMI failure. In the EDC test, a tensile stress in axial direction was applied and displacement-controlled loading was performed to keep the strain ratio of axial/hoop as a constant. The data of cladding deformation had been achieved in the range of strain ratio of 0, 0.25, 0.5 and 0.75 and pre-crack depth of 41–87 micrometers. Failures in hoop direction were observed in all the tested samples, and a general trend that higher strain ratio and deeper crack depth lead to lower failure limit in hoop direction could be seen. Different crack propagation mode was observed between recrystallized and stress relieved and cold worked samples, which might be due to the difference in microstructure caused by the final heat treatment at the fabrication of cladding.
Development of Probabilistic Risk Assessment Methodology of Decay Heat Removal Function Against Combination Hazard of Low Temperature and Snow for Sodium-Cooled Fast Reactors
Authors: Nishino Hiroyuki , Yamano Hidemasa , Kurisaka Kenichi
Published: October 2017

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions.

Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth.

To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure.

By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.
Assessment on Direct Debris Interaction at High Pressure Melt Ejection Scenario in Level 2 PSA
Authors: Hirokawa Naoki , Chitose Hiromasa , Hisamochi Kohei
Published: October 2017

Advanced Boiling Water Reactor (ABWR) has multiple safety features to prevent core damage and mitigate the accident progression. Failures of all injection systems to the reactor (ECCS, alternative injection systems, mobile injection systems) could cause a core melt accident and the Reactor Pressure Vessel (RPV) failure. If large amount of debris particles are deposited on the structures in the containment such as access tunnel, hatches, and Vacuum Breakers (V/Bs), it may result in containment failure or Suppression Pool (S/P) bypass due to the direct heating from the corium. This is defined as Direct Debris Interaction (DDI) in this paper.

The phenomenon is introduced in the containment event trees of internal events at power Level 2 Probabilistic Safety Assessment (PSA) for an ABWR and sensitivity analyses and source term analyses are performed in this paper. As a result, the contribution of DDI to total Containment Failure Frequency (CFF) is not significant for the ABWR even if the conservative DDI probability is assumed.
Radiation Shielding Analysis of Radioactive Material Product Container
Authors: Tian Yingnan , Long Lin , Zhang Puzhong , Gao Guiling , Mi Aijun , Li Zhuoran , Mao Yawei
Published: October 2017

For nuclear fuel reprocessing plants, some materials with high activity have been collected in the product container in the process of spent fuel reprocessing. In order to ensure the staff’s occupational radiation safety, an analysis of product container radiation shielding and source terms is crucial to determine the total activity limit and the activity limit range of different rays while reprocessing radioactive materials in the product container, so as to determine the working time limitation. A basic calculation result shows that generally alpha particles are not taken into consideration when calculating its occupational radiation effect for product container. In order to reduce the radiation exposure caused by neutron and gamma ray, radiation nuclides such as actinides and activated products which have a high neutron and gamma yield, and also the total nuclide amount should be controlled. In addition, the purification of nuclide emitting pure beta ray such as 90Sr, 90Y, 106Ru, 106Rh should be strictly supervised in order to restrict the beta ray exposure and its bremsstrahlung effect.
Numerical Simulation of Flow Distribution in the Lower Plenum of NHR200-II
Authors: Li Wei , Wang Dingqu , Jiang Yueyuan , Li Songyang , Guo Wenli , Xiong Wei , Li Weihua , Xie Heng , Li Xiaotian
Published: October 2017

The type-II 200MW nuclear heating reactor (NHR200-II) designed by Tsinghua University is a novel pressurized water reactor, whose main specifications are as follows, the heat power is 200 MW, the design pressure of major loop is 10 MPa, the inlet and outlet temperature of coolant are 230 °C and 278 °C. In this paper, a quarter three-dimensional model of the lower plenum of the reactor pressure vessel is set up for analysis. The flow velocity distribution and the pressure distribution on the core supporting structure are calculated by method of three-dimensional numerical simulation. The results show that the lower part of the core produce symmetric vortex due to the existence of support structure. The production of the symmetric vortex, to some extent, increases the instability of the flow. On the other hand, the existence of the vortex is good for uniformity of flow distribution in the outlet holes. The flow rate in the flow channel of support structure is lower at the center and larger in the margin of core inlet. The results show that the maximum of the velocity in the flow channel is 5% higher than the minimum one.
Derivation of Body Force Model Under Ocean Conditions
Authors: Rao Yuxian , Yu Lei
Published: October 2017

The floating nuclear power plant (FNPP) is recently received increasing attention in China. The effects of ship motions on the working fluid of marine-type reactor are not only a change of effective gravity, but also an appearance of additional forces, which requires further researches. This paper focuses on the derivation of body force model under 6 typical ocean conditions. The derived model consists of body force terms based on both three-dimensional and one-dimensional momentum equations and can be applied to the pipe with an arbitrary flow direction. For programming calculation, the integrations with respect to spatial and time variables are performed for the body force terms of one-dimensional momentum equations under 6 typical ocean conditions. For verification purpose, the deprived model is implemented into the RELAP5 code and a two-loop single phase natural circulation system is analyzed using the modified RELAP5 code under 6 typical ocean conditions. The calculation results show that the deprived model is reasonable.
A Study of Seismic Probability Safety Assessment for a Typical Three-Loop Pressurized Water Reactor Nuclear Power Plant
Authors: Wang Yuqing , Li Lingzhi , Wang Dongmei
Published: October 2017

An introduction of the background, purpose and the methodology of seismic Probability Safety Assessment (PSA) for the typical three-loop pressurized water reactor nuclear power plant are firstly given. Seismic hazard curves for the site of the plant are developed through Probabilistic Seismic Hazard Analysis (PSHA). With the input of site specific Uniform Hazard Response Spectrums (UHRSs), Soil-Structure Interaction analysis is completed using multiple time-history analysis method. After the development of the seismic equipment list, a screening process is completed with screening criteria determined on the basis of seismic hazard curves and a rough estimate of the seismic risk level. For the structures and components not screened out, seismic fragility parameters are calculated following the separation of variables approach. A seismic PSA model is developed afterwards, in which a seismic pre-tree and several seismic event trees as well as relating fault trees are built to model the accident sequence in seismic events. The quantification of the model is done on the software of RiskSpectrum and its additional code for hazard analysis, RiskSpectrum.HazardLite. According to the results, seismic induced Core Damage Frequency (CDF) of the plant is 2.69×10−6/reactor year, and the range of Peak Ground Acceleration (PGA) of seismic events from 0.30g to 0.45g has the most significant risk contribution. Seismic induced loss of offsite power, loss of emergency AC power and DC power are important accidents for the plant. Electrical components, auxiliary feeding water tank, diesel generators and ceiling of main control room of the plant are important contributors to seismic risk.
Discussion About Public Participation in Environmental Impact Assessment in China’s Nuclear Power Plant Project
Authors: Yuan Fang
Published: October 2017

Public participation systems in environmental impact assessment started late in our country. Relevant laws, regulations, and work protocols need to be further improved. In this study, extensive research was conducted on the public participation systems in the environmental impact assessment of foreign nuclear power plants. Analyze the current status of our public participation systems were drawn from legal aspect and the aspect of implementation. Together with case analysis, main problems of public participation systems in environmental impact assessment of China’s nuclear power plant were summarized from this study: (1) delayed information disclosure; (2) the scope of public participation need to be widened; (3) interactive platforms are required for convenient and efficient public participations instead of a single participation approach; (4) timely response to the platforms and more supervision over the participation systems are desired. Solutions to each problem are proposed to help develop relative regulations and the implementation of these regulations.
Double-Wall Natural Draft Heat Exchanger Design for Tritium Control in FHRs
Authors: Zhang Sheng , Shi Shanbin , Wu Xiao , Sun Xiaodong , Christensen Richard
Published: October 2017

Tritium control is potentially a critical issue for Fluoride salt-cooled High-temperature Reactors (FHRs) and Molten Salt Reactors (MSRs). Tritium production rate in these reactors can be significantly higher compared to that in Light Water Reactors (LWRs). Tritium is highly permeable at high temperatures through reactor structures, especially. Therefore, heat exchangers with large heat transfer areas in FHRs and MSRs provide practical paths for the tritium generated in the primary salt migrating into the surroundings, such as Natural Draft Heat Exchangers (NDHXs) in the direct reactor auxiliary cooling system (DRACS), which are proposed as a passive decay heat removal system for these reactors.

A double-wall heat exchanger design was proposed in the literature to significantly minimize the tritium release rate to the environment in FHRs. This unique shell and tube heat exchanger design adopts a three-fluid design concept and each of the heat exchanger tube consists of an inner tube and an outer tube. Each of these tube units forms three flow passages, i.e., the inner channel, annular channel, and outer channel. While this type of heat exchangers was proposed, few such heat exchangers have been designed in the literature, taking into account both heat and tritium mass transfer performance.

In this study, a one-dimensional heat and mass transfer model was developed to assist the design of a double-wall NDHX for FHRs. In this model, the molten salt and air flow through the inner and outer channels, respectively. A selected sweep gas acting as a tritium removal medium flows in the annular channel and takes tritium away to minimize tritium leakage to the air flowing in the outer channel. The heat transfer model was benchmarked against a Computational Fluid Dynamics (CFD) code, i.e., ANSYS Fluent. Good agreement was obtained between the model simulation and Fluent analysis. In addition, the heat and mass transfer models combined with non-dominated sorting in generic algorithms (NSGA) were applied to investigate a potential NDHX design in Advanced High-Temperature Reactor (AHTR), a pre-conceptual FHR design developed by the Oak Ridge National Laboratory. A double-wall NDHX design using inner and outer fluted tubes was therefore optimized and compared with a single-wall design in terms of performance and economics.
Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents
Authors: Ishikawa Jun , Sugiyama Tomoyuki , Maruyama Yu
Published: October 2017

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2.

In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.
Theoretical and Experimental Research on the Single-Phase and Incipient Boiling Heat Transfer of Liquid Metal Sodium
Authors: Zhang Dalin , Chen Jing , Song Ping , Wang Shibao , Wang Xin'an , Li Linfeng , Qiu Suizheng , Su G. H.
Published: October 2017

Sodium-cooled Fast Reactor (SFR) is the most promising reactor among the six Gen-IV nuclear reactor systems. The coolant of SFR is liquid metal which has very different heat transfer characteristics from conventional fluids. The study on heat transfer characteristics of the liquid metal sodium at single-phase and two-phase has not only great academic but also application meanings for SFRs design and safety analysis. Xi’an Jiaotong University (XJTU) performed comprehensive studies from the theoretical and experimental aspects.

In the theoretical studies, the main focuses are paid on the sodium single-phase heat transfer flowing in rod bundles, incipient boiling, two-phase thermal-hydraulic and critical heat flux (CHF). The experimental research on the flow and heat transfer characteristics of sodium in annuli is performed on XJTU single-phase sodium loop and two-phase boiling sodium loop, in which the single-phase friction and Nu correlations, the incipient boiling superheat (IBS) correlations, the two-phase friction multiplier factor and the heat transfer coefficient of sodium boiling are obtained. In this paper, the established theoretical models and the experimental studies for the single-phase heat transfer and the IBS in the annuli or the pin bundles are presented. The theoretical models are validated by the experimental data, and the comparison shows good agreement.
The Concept Study on the New Steam Cooling Reactor
Authors: Yan Xuesong , Zhang Xunchao , Zhang Yaling , Yang Yangyang , Yang Lei , Duan Wenshan
Published: October 2017

Coolant is an important part of the nuclear reactor, which can convert the fission energy into heat. At the same time, it can cool the core, and control the various components of the reactor in the allowable temperature. In this paper, gas-liquid two-phase flow is proposed as the coolant in the nuclear reactor. Based on the Monte Carlo method of particle transport program GMT program and LITAC program, the neutron performance, thermal hydraulic of nuclear reactor were calculated and analyzed. The operation of the new reactor will not be refuel for decades. The gas-liquid two-phase flow coolant can be used as one of the existing alternative coolants.
Development of Numerical Estimation Method Using Spatial Connection Methodology for Thermal Striping in Upper Plenum of Reactor Vessel of an Advanced Loop-Type Sodium-Cooled Fast Reactor in Japan
Authors: Tanaka Masaaki , Murakami Satoshi
Published: October 2017

Thermal striping on the core instrumentation plate (CIP) at the bottom of the upper internal structure (UIS) of an advanced loop-type sodium-cooled fast reactor in Japan (Advanced-SFR) has been numerically investigated. At the top of the core below the CIP, the sodium at high temperature flows out from the fuel subassemblies (FSs) and the sodium at low temperature flows out from the primary control rod (PCR) and backup control rod (BCR) channels, and also the radial blanket fuel subassemblies (RBFSs) at the outer side of the core. In order to predict the thermal striping on the CIP caused by mixing fluids at different temperatures from the FSs, the PCR and the BCR channels, and the RBFSs, a numerical estimation method using a spatial connection methodology between the upper plenum analysis and the local area analysis for the target area has been developed. By using the connection methodology, the numerical simulation considering the influence of the transversal flow in the UIS and the external flow around the UIS in the upper plenum can be performed to improve the accuracy of the estimation results. In this paper, the outline of the spatial connection methodology including data transfer technique from the upper plenum analysis to the local area analysis was described. As a validation process, numerical simulation of the water experiment using the test apparatus named TAFUT which was 1/3-scaled 1/6 partial model of the upper plenum of the Advanced-SFR was performed to confirm applicability of the spatial connection methodology to a practical thermal striping problem. The numerical result of temperature distribution was compared with the measured result in TAFUT experiment. Additionally, mesh sensitivity of the local area analysis model to the numerical results was indicated by using a small and a large area models in order to suggest an appropriate local area analysis model.
The Application of Task Analysis and Task Support Verification in HPR1000 Main Control Room Human System Interface Design
Authors: Shiguang Deng , Xiaoyang Ming , Yuan Gao
Published: October 2017

As the brain of Nuclear Power Plant (NPP), Main Control Room (MCR) is an integrated system which comprises varieties of Human System Interfaces (HSIs), procedures, operators and environment. HSI is the medium that the operators monitor and control the Nuclear Power Plant. To improve the operating performance and avoid human errors, Human Factor Engineering (HFE) is required to be implemented in the new plant HSI design for plant safety.

HPR1000 has followed the NUREG-0711 as the guidelines of HFE program, in which 12 elements are included. Task Analysis (TA) and Human Factor Verification & Validation (HF V&V) are two important elements for the HSI design. For task analysis, the design engineers could analyze the human actions (HAs) to identify the tasks that the operators must perform to accomplish the plant functions. Task analysis identifies not only the specific tasks needed to accomplish HAs but also the information, and controls required to complete those tasks which are considered as important input for the MCR HSI design.

HF V&V is one of the final stages of the Human Factors Services as included in NUREG-0711. The HF V&V objective is to evaluate that the HSI design conforms to HFE design principles and that the adequacy of the interfaces enable plant personnel to successfully perform their tasks and other operational goals, assuring plant safety. The V&V element consists of four major activities: Sampling of Operational Conditions, Design Verification, Integrated System Validation, and HED Resolution. For different targets, Design Verification could be separated into two kinds of activities: HSI Task Support Verification (TSV) and HFE Design Verification (DV). TSV is an evaluation to verify that the HSI supports the requirements of personnel tasks, as defined by task analyses. The HAs, controls and information identified in task analysis could be used as the basic input for HSI TSV.

To provide adequate input for the HSI design and TSV, the operational conditions of task analysis should be sampled as many as possible to cover all the typical conditions, which include normal conditions, transients, Design Based Accidents (DBA), Beyond Design Based accidents (BDBA), even severe accidents. The sampled conditions for task analysis would be available for the task support verification of the related HSIs.

In conclusion, the task analysis could be used as the design input for HSI, while the task support verification could review the designed HSI with the output information from TA. This paper addresses the research and engineering implementation methodologies for task analysis and HSI task support verification. As the first reactor of HPR1000, HFE program have been carried out in Fuqing 5&6 MCR from preliminary design to final test. The Fuqing 5&6 TA and TSV achievements for MCR HSIs design are summarized in the paper.
Data Flow Based Cyber Security Defense-in-Depth Model of I&C System for Nuclear Power Plants
Authors: Hu Bing , Zhang Longqiang , Guo Zhiwu , Li Youran , Sun Wei , Zhou Liang , Tian Yong
Published: October 2017

With the introduction of digital instrumentation system, the cyber security threat to nuclear power plants is becoming more and more serious. The existing cyber security standards of nuclear power plants still need to be improved, and the technology practice of defensive strategies is lacking all over the world. In this paper, based on the comparison of domestic and foreign regulations and standards, combined with the technical practice of I&C system overall plan, a defense-in-depth model based on data flow is proposed. The overall technical requirements, hierarchy, network model, cyber security basic requirements, cyber security interface and protection of digital assets are introduced, the application of the model and the direction of research on cyber security of nuclear power plant are prospected.
Measurements of Two-Phase Pressure Drop in a Simulated Debris Bed
Authors: Nava Rivera Milka Hebi , Ito Daisuke , Saito Yasushi , Aoyagi Mitsuhiro , Kamiyama Kenji , Suzuki Tohru
Published: October 2017

Two-phase flow through porous media should be well understood to develop a severe accident analysis code not only for light water reactor but also sodium cooled fast reactor (SFR). When a core disruptive accident occurs in SFR, the fuel inside the core may become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. Thus, as first step, present work focuses on the characteristics of pressure drop in single- and two-phase flows in different porous media conditions (porous size, liquid and gas flow velocity). In addition, in order to construct an experimental database, the measured pressure drop under different conditions was compared with existing correlations.
Application of Fluid-Structure Interaction Analysis in the Design of Upper Water Tank Structure of HPR1000 Outer Containment
Authors: Di Yao , Lijian Cai , Jian Meng , Jintao Zhao
Published: October 2017

Based on the basic principle of fluid-structure interaction, this paper make a finite element analysis of seismic on upper water tank of HPR1000 outer containment by CEL method in ABAQUS software. Firstly, structure is simulated the by Lagrange grid and the water in upper water tank by Eulerian grid; secondly, coupling contact between water and structure is defined; finally, the calculation results are got by running an explicit dynamic solver to makes a time history analysis of fluid-structure interaction finite element model under the seismic, and the results will be used in the structure design of outer containment and upper water tank.
Study on Radiation Effects on AMP/SiO2 Adsorbent for Separation of Cesium From High Level Radioactive Waste
Authors: Wu Yan , Zhang Xiaoxia , Wei Yuezhou
Published: October 2017

A silica-based ammonium molybdophosphate (AMP/SiO2) was successfully applied to remove Cs(I) from radioactive waste water in our previous study. The structure characterization of irradiated AMP/SiO2 was carried out by X-ray diffraction (XRD). The effect of the initial concentration of metal ions and acidity concentration for removal Cs(I) by irradiated AMP/SiO2 were investigated. No obvious difference was observed for AMP/SiO2 with 300 kGy absorbed dose from the XRD patterns, except the slightly shift of (110) peak toward a small angle. Relatively large distribution coefficient (8×103cm3/g) of Cs(I) was obtained in 3M HNO3 by irradiated AMP/SiO2. The maximum adsorption capacity of AMP/SiO2 irradiated at 300 kGy was estimated as 22.4 mg/g. The loss of molybdenum from AMP/SiO2 in 3 M HNO3 with 200 kGy absorbed dose is less than 5%.
Loading Scheme Research of Small Modular Sodium Cooled Reactor in a Factory
Authors: Xiao Hu , Xiao Liang Chen
Published: October 2017

An innovative small modular sodium cooled fast reactor called S1 is designed by China Institute of Atomic Energy (CIAE). As an encapsulated nuclear source with thermal power of 3MW, S1 is characterized by small volume, light weight, high safety and sound reliability. The S1 systems adopt modularization, by which the core will be loaded in a factory and filled with sodium, then shipped to the assembled onsite, thus the construction time for S1 can be substantially reduced.

In this paper, the Monte Carlo code MCNP is used to calculate the loading scheme of S1. Considering factory’s special conditions for loading and active chemical property of sodium, a special loading pattern is adopted for S1 — loading fuel first, then sodium filling.

Ensure that the neutron source and detectors are well matched during loading, and detector counting rate is no less than 2cps when only neutron source but no fuel exists in the core. Three positions where the 252Cf neutron source is placed are studied in this paper: (1) at the center of the core; (2) at the middle of outside core plane; (3) at the bottom of outside the reactor vessel. Through MCNP simulation calculations and comparison of large resulting data, it finds the neutron source should be reasonably placed at the bottom of the reactor vessel where 252Cf strength is 105 s−1 neutrons, and the ex-core detectors are distributed symmetrically at the center of outside core plane; the most befitting moderator material of detector surface is methacrylate-C5H8O3. In this paper, 1/N extrapolation method is used during loading and kinds of loading schemes have been studied with reference to the principles of China experimental fast reactor (CEFR) and regulations of relevant research reactors, and 5-batch loading scheme is finally chosen as the optimal loading scheme. S1 is prepared for sodium filling at 250 °C. It shows that neutron flux variation of core can be more reliably monitored when the ex-core detectors are placed about 120cm away from the center core through MCNP simulation calculation. Such arrangement can also meet the monitoring requirements for loading and sodium filling.
Sensitivity Analysis on the Detailed Plant Operational State for Shutdown PSA
Authors: Shiota Daichi , Hirokawa Naoki , Hisamochi Kohei
Published: October 2017

Probabilistic Safety Assessment (PSA) for the plant shutdown state of nuclear power plants is performed based on Plant Operational States (POSs) which depend on periodic inspections and maintenance. For a typical BWR, the shutdown POSs are defined in consideration of various plant conditions: boundaries of Primary Containment Vessel (PCV)/ Reactor Pressure Vessel (RPV), reactor water level, availability of mitigation systems and temperature/pressure of primary coolant. The shutdown POSs for a BWR plant may involve conservative simplifications. One of the simplifications is treatment of PCV/RPV heads conditions during the POSs involving their opening/closing operations. Usually the PCV/RPV heads conditions are assumed to be the most onerous in terms of accident sequence modelling on both Level 1 and 2 PSA.

In this study, a sensitivity analysis is performed to understand the significance of the conservatism by refining the POSs which involve opening/closing operations of PCV/RPV heads.
Preliminary Study on the Identification of DECs Without Significant Fuel Degradation in New NPP Designs
Authors: Yang Zhiyi , Chong Yimin , Li Chun , Zhang Jiajia
Published: October 2017

New nuclear safety objectives and principles are being studied in main nuclear power countries and organizations after Fukushima Dai-ichi nuclear accident, to further improve the safety level of nuclear power plants (NPPs). Based on International Atomic Energy Agency (IAEA) Specific Safety Requirements (No.SSR-2/1), “Safety of Nuclear Power Plants: Design” (HAF102-2016) is issued in China. The concept “design extension condition (DEC)” is put forward, which is intend to enhance the plant’s capability to withstand accidents that are more severe than Design Basis Accidents (DBA). DEC could include conditions without significant fuel degradation (DEC-A in this paper) and conditions with core melting (DEC-B in this paper), e.g. severe accident. In this paper, the DEC-A and its application was discussed preliminarily, firstly, the development and connotation was introduced, then the identification of DEC-A, and the safety analysis principles of DEC-A were mainly described. This study may play a valuable role for implementation of new nuclear safety requirements in China.
Management of Geohazards Caused by the Impact of New Infrastructure Projects on Common Right of Way (ROW)
Authors: Aristizábal Ceballos Jaime Hernán , Chaves Agudelo Julián , Castellanos Barandica Luis C.
Published: October 2017

Integrity Management on Cenit common Right-of-Way (ROW), contemplate a preventive vision respect the impact that new infrastructure projects can cause in hydrocarbon transportation systems directly and indirectly. By including the management of Climate and External Forces Threat from this preventive approach have been identified mitigation and/or maintenance actions of which must be considered in the arrangements of common right-of-way with external entities responsible for the planning, construction and operation of these projects. This document presents the management strategy that was built an implemented largely, and the exposed examples are intended to show the importance of incorporates these learned lessons in the management of geotechnical assurance of the ROW.
Case of Geotechnical Instrumentation of Pipelines in Unstable Zones: Real Time Readings and its Development in Uncommunicated Zones
Authors: Robles Robles Dimas , Castillo Martínez Einstein
Published: October 2017

Oil pipelines and gas pipelines usually go through geotechnically unstable areas for different reasons. These can go from situations related to the engineering stage (trace), to environmental and social aspects during the construction process. Due to these aspects, the ducts go through geotechnically undesirable areas. Usually, the geotechnical instabilities, according to the kind of movement, are low speed (cm/year), medium (m/year) and very quick processes that generate movements of tens to hundreds of meters per day. Most of Mass Removal Phenomenon (MRF) are triggered by rain and/or earthquakes and are translated into land movements which at the same time involve, occasionally, important deformations in pipelines or its breaking, depending on the movement speed and the possibility of making works before the pipeline breaking. To get to know the pipeline tensional state from the beginning of the pipeline operation, in this unstable zones, is an essential task, which depends on the early identification of the said land movements and the possibility to do measurements on the pipelines using tools such as In-line inspection running (ILI) or the installation of strain gauges. This situation makes the task of monitoring in unstable zones a vital one. The current paper is based on a breaking pipeline case due to soil movement, “monitored by inclinometers”, with the purpose to show the importance of a geotechnical and mechanical instrumentation that offers useful results. The instrumentation allows to model the interaction soil-pipeline to accomplish relevant tasks, that avoid the pipeline breaking and at the same time allow to stablish deformation thresholds of soil or pipeline, which will become early warnings to avoid breakings. Furthermore, the soil and pipeline’s deformation thresholds are documented, based on a system transport by pipelines (STP) breaking cases, to stablish threat classifications to a specific pipeline.

The called instrument reading in real time implies: detection, measurement and data broadcasting that allows the user to have daily records of the movements or required associated variables, with no need to depend on other communication systems that might be inexistent in some areas. This paper also shows the development and operation of a monitoring station that includes: inclinometers, piezometers, strain gauges and rain gauges, among others. These broadcast their data to a server that the user has access to, from any place with a Wi-Fi network, here the user will be able to display information from each one of the instruments, emphasizing the measured variables or magnitudes (displacement, water level, micro strain mm/day) into graphics. The station has a limitation over battery length of 6 months, when it’s problematic to install a recharge solar cell system.
Best Practices for the Strain Relief Excavation by Geotechnical Movements in Ocensa
Authors: Ramos Adalgiza , Ocampo Andrés
Published: October 2017

This article aims to describe the best industry practices followed by the line maintenance area of the Ocensa pipeline, which can be divided into 4 large steps that follow a PHVA cycle of the activity, which includes relief planning, the execution of the same, the verification and geotechnical monitoring, closure and feedback.
Frequency shifting reduces but does not eliminate acoustic interference between echolocating bats: A theoretical analysis
Authors: Perkins Melinda L. , Frank Hannah K. , Pauly John M. , Hadly Elizabeth A.
Published: October 2017

Bats have been observed to shift the frequency of their echolocation calls in the presence of other echolocating bats, ostensibly as a way to reduce acoustic interference. Few studies, however, have examined the theoretical efficacy of such jamming avoidance responses. The present study uses the wideband ambiguity function to analyze the effects of acoustic interference from conspecifics and congeneric heterospecifics on the target acquisition ability of Myotis californicus and Myotis yumanensis, specifically whether unilateral or bilateral frequency shifts reduce the effects of such interference. Model results suggest that in conspecific interactions, M. yumanensis recovers its target acquisition ability more completely and with less absolute frequency shift than does M. californicus, but that alternative methods of jamming avoidance may be easier to implement. The optimal strategy for reducing heterospecific interference is for M. californicus to downshift its call and M. yumanensis to upshift its call, which exaggerates a preexisting difference in mean frequency between the calls of the two species. Further empirical research would elucidate whether these species do in practice actively employ frequency shifting or other means for jamming avoidance, as well as illuminate the role of acoustic interference in niche partitioning.
Effects of ejective stops on preceding vowel duration
Authors: Beguš Gašper
Published: October 2017

One of the most widely studied observations in linguistic phonetics is that, all else being equal, vowels are longer before voiced than before voiceless obstruents. The causes of this phonetic generalization are, however, poorly understood and several competing explanations have been proposed. No studies have so far measured vowel duration before stops with yet another laryngeal feature: ejectives. This study fills this gap and presents results from an experiment that measures vowel duration before stops with all three laryngeal features in Georgian and models effects of both closure and voice onset time (VOT) on preceding vowel duration at the same time. The results show that vowels have significantly different durations before all three series of stops, voiced, ejective, and voiceless aspirated, even when closure and VOT durations are controlled for. The results also suggest that closure and VOT durations are inversely correlated with preceding vowel duration. These results combined bear several implications for the discussion of causes of vowel duration differences: the data support the hypotheses that claim that laryngeal gestures, temporal compensation, and closure velocity affect vowel duration. Some explanations, especially perceptual and airflow expenditure explanations, are considerably weakened by the results.
Temporal variation in pelagic food chain length in response to environmental change
Authors: Ruiz-Cooley Rocio I. , Gerrodette Tim , Fiedler Paul C. , Chivers Susan J. , Danil Kerri , Ballance Lisa T.
Published: October 2017

Extreme oceanographic conditions are associated with shorter food chain length in the California Current ecosystem.
Analysis of China's one-child policy sparksuproar
Authors: Hvistendahl Mara
Published: October 2017

Colleagues call demographer's findings flawed and irresponsible.
Comment on “Persistent effects of pre-Columbian plant domestication on Amazonian forest composition”
Authors: McMichael Crystal H. , Feeley Kenneth J. , Dick Christopher W. , Piperno Dolores R. , Bush Mark B.
Published: October 2017

Levis
et al
. (Research Articles, 3 March 2017, p. 925) concluded that pre-Columbian tree domestication has shaped present-day Amazonian forest composition. The study, however, downplays five centuries of human influence following European arrival to the Americas. We show that the effects of post-Columbian activities in Amazonia are likely to have played a larger role than pre-Columbian ones in shaping the observed floristic patterns.