bluuebunny/crossref_metadata_2025 huggingface dataset:
Flow Field Analysis of the Reactor Coolant Pump Internal Cooling Circulation
Authors: Qin Jie , Xu Qingmu , Yuan Junkai , Cai Kun
Published: October 2017

Reactor coolant pump (RCP) is one of the most critical devices in third generation of pressurized water reactor nuclear power plant. EMD shield pump and KSB wet winding pump are two representative kinds of RCPs without complex shaft seal system. Due to cancellation of shaft seal system, the entire rotors (including the flywheel) are immersed in the coolant. The losses in RCPs take one third of the total power including rotation loss caused by rotor in the water, electromagnetic loss in the shielding sleeve,the heat transferred through high temperature coolant, and heat generated by bearing.Because of the losses listed above, bearing and winding are heated,and the losses make temperature rise. in order to ensure that the motor is working properly at low temperatures, the company EMD and KSB design the RCP internal cooling circulation which brings the heat out to ensure the normal operation of the RCPs.

The RCP internal cooling circulation includes inlet flow area, auxiliary impeller, thrust bearing, the lower flywheel, motor can, upper radial bearing, upper flywheel, outlet flow area, and external heat exchanger,etc. Flow characteristics in every flow path determine the flow distribution and heat transfer, and the flow distribution determines whether the cooling performance of RCP internal cooling circulation meets the requirements. In order to control operating temperature of motor and bearing, and to optimize heat transfer, adjusting the size of flow area and changing the flow characteristics arecritical. flow field and temperature field in RCP internal cooling circulation need overall analysis.

Flow distribution can be obtained theoretically through the calculation of an overall three-dimensional model.But on the one hand, the calculation time is long due to a complex three-dimensional model with a large quantity of grids, on the other hand, it is easier to casue errors in local processing and the errors are difficult to find or correct. For rapid analysis and optimization of flow and heat transfer in RCP internal cooling circulation, ensure the motor winding and bearing operate at an appropriate temperature, the local characteristics of RCP internal cooling circulation are studied, one-dimensionalanalysis method of RCP internal cooling circulation is developed. This one-dimensional analysis method can be used to predict the flow distribution of each part of RCP internal cooling circulation according to change of the channel geometry parameters, key dimensions, boundary conditions and rotor speed. The geometric parameters are optimized by analyzing the flow distribution, and the purpose of design guidance are achieved.
Leakage Flow Investigations on a Through-Wall Crack Related to Thermal Fatigue of Nuclear Power Plant Piping
Authors: Schmid Stefan , Kulenovic Rudi , Laurien Eckart
Published: October 2017

For the validation of empirical models to calculate leakage flow rates in through-wall cracks of piping, reliable experimental data are essential. In this context, the Leakage Flow (LF) test rig was built up at the IKE for measurements of leakage flow rates with reduced pressure (maximum 1 MPA) and temperature (maximum 170 °C) compared to real plant conditions. The design of the test rig enables experimental investigations of through-wall cracks with different geometries and orientations by means of circular blank sheets with integrated cracks which are installed in the tubular test section of the test rig.

In the paper, the experimental LF set-up and used measurement techniques are explained in detail. Furthermore, first leakage flow measurement results for one through-wall crack geometry and different imposed fluid pressures at ambient temperature conditions are presented and discussed. As an additional aspect the experimental data are used for the determination of the flow resistance of the investigated leak channel. Finally, the experimental results are compared with numerical results of WinLeck calculations to prove specifically in WinLeck implemented numerical models.
Experimental Investigation on the Pipe Leakage
Authors: Cai Benan , Zhang Qi , Weng Yu , Gu Hongfang , Wang Haijun
Published: October 2017

Pipelines are widely used in many fields including power industry, petroleum system etc. Pipelines such as the surge line and main pipe are easily subjected to thermal stratification as a result of the non-uniform temperature distribution in the nuclear power plants. Furthermore, pipelines can suffer from thermal fatigue in virtue of long-term uneven stress distribution. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen because of the existence of pipeline crack. The thermal pipeline crack leakage mainly appears in the region with stress concentration. As the pipe system is always covered with thermal insulation layer in the actual nuclear power plants, it is hard for workers to observe pipeline leak, which can have a bad effect on the normal operation. Since the temperature and humidity close to the pipe crack due to leakage can change compared to the normal operation, we can infer from the temperature and humidity changes that the pipe leakage occurs. Based on this idea, the temperature and humidity near the crack of the pipe need to be measured to detect the leakage fields. As the fluids with high pressure and high temperature flow in the pipe system in an actual nuclear power plant, the pipe leakage experiment was performed in the high pressure and high temperature condition. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The temperature and humidity variations were monitored in the experiment with temperature and humidity probes which have the advantage of responding to the change of temperature and humidity sensitively. The data collection program was mainly written based on the LABVIEW platform. The collecting time step was set 1s. As the measuring position and leakage flux are two key factors for the pipe leakage, the experiment was carried out with different measuring positions and leakage fluxes conditions. The experimental results showed that the leak flux had an important influence on the temperature and humidity near the pipe crack. The temperature and humidity started to change in a very short time with large leak flux. At the same time, the velocity of the temperature and humidity change was high with large leak flux. When the pipe leakage occurred in the location near the temperature and humidity probe, the temperature and humidity responded quickly and the velocity of temperature and humidity change was large. The experiment data can be used for the prediction of the pipe leakage in the nuclear power plants.
Neutronic Analyses for CFETR With Modular Helium Cooled Lithium Ceramic Blanket
Authors: Xu Kun , Ye Minyou , Song Yuntao , Lei Mingzhun , Mao Shifeng
Published: October 2017

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by national integration design group for magnetic confinement fusion reactor of China to bridge the R&D gaps between ITER and DEMO. Since the launch of CFETR conceptual design, a modular helium cooled lithium ceramic blanket concept had been under development by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences, to complete CFETR in demonstrating its fusion energy production ability, tritium self-sufficiency and the remote maintenance strategy. To validate the feasibility, the neutronic analyses for CFETR with this modular helium cooled lithium ceramic blanket were performed. The 1-D neutronic study for CFETR was done in the first place to give a preliminary and quick demonstration of the overall neutronic performance. Meanwhile, the neutronic analyses for a single standard helium cooled lithium ceramic blanket module were done in several times to give more insight for the material and geometry parameters of intra-module structures. Therefore, the principles for neutronic design and the module level optimized parameters were produced, based on which the design of practical blanket modules planted in tokamak vacuum vessel was completed. In the end, the 3-D neutronic analysis for CFETR was done utilizing the MCNP code, in which the 11.25 degree sector model (consist of blanket modules, manifold, support plate, shield, divertor, vacuum vessel, thermal shield and TF coils) was generated with the McCad automated conversion tool from the reference CAD model for analysis, the bi-dimensional (radial and poloidal) neutron source map was plugged via general source definition card to stimulate the D-T fusion neutrons. The concerned neutronics parameters of CFETR, mainly including the tritium breeding ratio to characterize tritium self-sufficiency, the energy multiplication factor to characterize power generation, as well as, the inboard mid-plane radial profiles of neutron flux densities, helium production rate, displacement damage rate and the energy deposition to characterize the shielding performance, were produced.

In principle, the neutronics performance of CFETR with modular helium cooled lithium ceramic blanket is promising. The tritium breeding capability meets the design target and, by referring to that for ITER and the EU DEMO fusion power plant, the inboard mid-plane shielding is effective to fulfill the radiation design requirement of the superconducting TF-coils, resulting in a compulsory warm-up time interval of ∼2 FPY for TF-coils. The nuclear heating loads to other CFETR components were generated. As an outcome of this work, the applicability of McCad on CFETR neutronic modeling is demonstrated.
Long-Term Cooling Strategy for the Primary Containment Vessel of the Kashiwazaki-Kariwa Nuclear Power Station in a Severe Accident
Authors: Kawaguchi Satoshi , Mizuno Satoshi , Oyama Yoshihiro
Published: October 2017

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident.

If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower.

TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout).

This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment.

Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied.

(1) RPV heat removal by restoring the RHR system

(2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger

(3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers

(4) RPV heat removal using the clean up water system (CUW)

By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.
Reinforcement of Emergency Power Supply System for Next Generation SFR
Authors: Ishikawa Nobuyuki , Chikazawa Yoshitaka , Sato Daisuke , Ikari Risako
Published: October 2017

In this paper, the emergency power supply system for next generation sodium fast reactor (SFR) is investigated to target the latest system design with such as optimized equipment for measure of loss of heat removal system (LOHRS) events, taking into account for strengthening diversity and multiplicity. In addition to diversity and multiplicity, optimization in terms of amount of equipment is also taken into account. The emergency power supply system is consist of emergency generators for design based accident (DBA) and emergency generators to cope with the design extension condition (DEC) in order to power the additional cooling equipment of fuel handling system and reactor core cooling system as a measure for LOHRS event. Base on the load capacity and system operation, the configuration of emergency power supply system was examined aiming optimization of amount of equipment in accordance with some shared use taking into account the reliability. As a result, the emergency power supply system with 4 fixed-type gas turbine generators, 4 portable-type diesel generators and batteries was established.
Assessment of Choking Flow Models in RELAP5 for Subcooled Choking Flow Through a Small Axial Crack of a Steam Generator Tube
Authors: Brown Mark A. , Nguyen Hung , Revankar Shripad T. , Riznic Jovica
Published: October 2017

Choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant (NPP), but also to everyday operation. Current NPP steam generators operate on the leak-before-break approach. The ability to predict and estimate a leak rate through a steam generator tube crack is an important safety parameter. Knowledge of the maximum flow rate through a crack in the steam generator tube allows the coolant inventory to be designed accordingly while limiting losses during loss of coolant accidents. Here an assessment of the choking flow models in thermal-hydraulics code RELAP5/MOD3.3 is performed and its suitability to predict choking flow rates through small axial cracks of the steam generator tubes is evaluated based on previously collected experimental data. Three sets of the data were studied in this work which corresponds to steam generator tube crack sample 1, 2, and 3. Each sample has a wall thickness, channel length (L), of 1.285 mm to 1.3 mm. Exit areas of these samples are 5.22 mm2, 9.05 mm2, and 1.72 mm2 respectively. Samples 1 and 2 have the same flow channel length to hydraulics diameter ratio (L/D) of 2.9 whereas sample 3 has a L/D of 6.5. A pressure differential of 6.8 MPa was applied across the samples with a range of subcooling from 5 °C to 60 °C. Flow rates through these samples were modeled using the thermal-hydraulic system code RELAP5/MOD3.3. Simulation’s results are compared to experimental values and modeling techniques are discussed. It is found that both the Henry-Fauske (H-F) and Ransom-Trapp (R-T) models better predict choking mass flux for longer channels. As the channel length decreases both models’ predictions diverge from each other. While RELAP5/MOD3.3 has been shown to predict choking flow in large scale geometries, further investigation of data sets need to be done to determine if it is suited well for small channel lengths.
Physics Design of Epi-Thermal Neutron Beam for BNCT Based on C-MNSR
Authors: Dan Peng , Xiaobo Wu , Jin Lu , Qian Hao , Jingyan Hong , Yiguo Li
Published: October 2017

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) (235U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT.

The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 109n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 107n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10−11Gy · cm2 · n−1 and gamma dose contamination (Dγ/φepi) 2.18× 10−14 Gy·cm2·n−1.
Research on Fatigue Monitoring System Development for Nuclear Power Plant
Authors: Chuanli Zhao , Yinqiang Chen , Feng Xu , Yuanyuan Ji , Jiangguo Wang
Published: October 2017

Fatigue is the major aging mechanism for pipelines and components that are subject to transient during plant operation. In order to avoid abrupt accidents caused by fatigue failure, it is necessary to adopt corresponding techniques to monitor components condition due to fatigue damage. Experience shows that on-line fatigue monitoring system is the most effective way to monitor fatigue damage status of pressure-retaining boundary subjected to operational loads in real-time for NPPs. Several commercial fatigue monitoring systems have been put into application in the nuclear power plants. In this paper, the difference of the international fatigue monitoring systems were compared. As the key technologies of fatigue monitoring system development, the processing of operational load data and the stress influencing function development at monitored locations were researched in the paper. Furthermore, in order to study on the methodologies of the key issues of the system development, the experimental facility was established to improve the knowledge on fatigue monitoring system development for two kinds of input data modules, and the reasonability of the different methods in the study was proved reciprocally. Finally, the on-line fatigue monitoring system for pipelines was established to evaluate actual fatigue damage status.
Experimental Study of Performance of Cold Crucible Induction Melter in Melting Corium Simulant
Authors: Sun Caixin , Chen Ying , Chen Lian , Chang Huajian , Pei Jie , Tian Fang , Xu Xinchen
Published: October 2017

Cold crucible induction melter (CCIM) technology has been widely used in melting of non-metallic refractory materials with a high melting point and a requirement of high purity. This paper discusses the application of the CCIM technology in melting of reactor corium (simulated as zirconium oxide in this experiment).

In this experiment, the power supply is set at constant power control mode, with a design frequency of ∼100 kHz and a total output power of ∼ 400 kW. The cold crucible consists of copper tubes and a stainless bottom plate which has a dimension of 220 mm in diameter and 420 mm in height. The crucible is charged with zirconium oxide (ZrO2) powder and surrounded by a set of water-cooling induction coils which is designed shorter than the crucible in order to reduce energy loss in the water-cooling bottom plate. The top of the crucible and the coils are at the same height. A Zirconium ring is used to initiate heating process. ∼ kg ZrO2 powder is used for each experiment. Electrical variables of the induction melter and temperature of the melt are detected and recorded in the entire heating process.

For the result, almost 3/4 of the ZrO2 powder is melt and a thin crust (∼ 2–3 mm) is generated between the melt and the crucible. To be specific, the ZrO2 powder above the bottom of the induction coils is melt, but others are not. In the melting process, the maximum surface temperature of the melt has reached 2230°C. Furthermore, for better understanding of the relationship between the melting process and the changing process of the electrical variables, an equivalent circuit of the heating system is established and analyzed. It can be concluded that some electrical variables such as the quality factor (Q) of the induction coils and the practical frequency can reflect melting status. The conclusion is confirmed by the records transmitted by sensors in the melting process. By comparison between the recorded data and the melting status, it is clear that the Q value can be accurately used in indicating the melting status due to its obvious change in the entire melting process. As the melting pool becoming larger, the Q value tends to be smaller. In contrast, the practical frequency becomes higher as the melt becoming larger. Furthermore, both of them are not influenced by the output power. In addition, the output current and the output voltage also can be used to describe the melting status. In specific, a larger melting pool corresponds to a lower output current and a higher output voltage.
The Integrated Management System for New Nuclear Power Projects: Requirements of IAEA Safety Standards
Authors: Xiaoping Li
Published: October 2017

The number of interested countries to build its first nuclear power plant remains high after the infamous Fukushima Daiichi Nuclear Accident. The development of the national infrastructure is one of the major undertakings for countries to construct nuclear power projects especially for the first nuclear power project. Management is one of the 19 key infrastructure issues and vital areas for the famous IAEA Integrated Nuclear Infrastructure Review mission. IAEA develops the concept of the Integrated Management System to ensure that the Safety, Health, Security, Safeguards, Quality, Economic and Environment issues are considered together in an organization, and to highlight the importance of safety in decision making. The IAEA Safety Standards establish requirements and provide guidance for applying a management system for facilities and activities. These IAEA Safety Standards mainly include Leadership and Management for Safety, The Management Systems for Facilities and Activities, and its applications addressing the different life cycle phases of a nuclear facility from design through operation to decommissioning and waste management. The author analyzed the current worldwide situation of the new nuclear power projects, briefly introduced the background of the infrastructure development of the nuclear power programme and INIR mission, and highlighted the IMS and IAEA safety standards on IMS and its applications with the practices from the ongoing new nuclear power projects in this paper.
Conceptual Design for CHP System for Class III Power
Authors: Sarvendran Vajran , Patel Heli , Harvel Glenn
Published: October 2017

Nuclear power plants (NPPs) typically have four classes of electrical supply systems to provide redundant and resilient power to ensure safe operation, cooling, and shutdown of NPPs. Class III emergency power traditionally uses standby diesel generators to fulfill equipment electrical demands. This requires a reliable source of diesel energy and reliance on one particular kind of fuel source for backup power.

During the North American blackout of 2003, hospitals had to run on backup generators that were mostly diesel fired. A prolonged blackout could have negatively affected hospital infrastructure and patient health. With current diesel generators, capable of supporting critical systems for up to 72 hours (for most NPPs), a prolonged loss of power or transient could similarly also have adverse effects on NPPs.

A Combined Heat and Power (CHP) system is a power generation system that generates electricity, in addition to concurrently having heating and cooling capabilities. Often, gas or steam turbines are used to generate electricity. CHP heating and cooling capabilities can be met via absorption coolers and heat pumps. Low-grade exhaust heat from turbines could be used for the absorption processes. Absorption coolers/heaters produce cold/warm fluid using a heat source via the vapor compression cycle, taking advantage of high affinity fluids. The absorbent allows for the refrigerant to boil at lower operating conditions, which allows for heat transfer.

CHPs’ that are capable of accepting natural gas as a form of fuel will increase resiliency of NPP power supply. This ensures a reduced risk of running out of fuel during prolonged transients due to continual supply via pipelines already in place.

The objective of this work is to generate a conceptual design of a multi-fuel source CHP system that is capable of at least accepting natural gas as an alternative to Class III diesel generators for a NPP. The system will be capable of supplying Class III power to Small Modular Reactor (SMRs), as well as commercial NPPs, such as CANDU, PWR, and BWR types.

Current state of the art Class III backup power systems, CHP systems, CHP thermodynamic cycles, multiple compatible fuels, as well as absorber heaters and coolers have been investigated. Ranking systems were used to determine the top three designs. The parameters included turbine type, efficiency, flow rates, operating temperature/pressure, fuel type, fuel to energy ratio, fuel availability, absorber chiller/heater fluid efficiency, operating temperature and thermal conductivity.

Based on the ranking system parameters, a thermodynamic model including mass, energy and entropy balance of an 8 MWe CHP with heating and cooling capability between the ranges of 15°C to 27°C was conceptually designed.
Design of an Assessment Tool for Site Evaluation of Nuclear Power Plants
Authors: Arthurs Cole , Doxtator Andrew , Mann David , Mann Shawn , Harvel Glenn
Published: October 2017

With the purpose of providing preliminary assessments of specific locations in their ability to support a nuclear power plant using a quantitative method, a tool has been developed to accept input from users and provide an output of an evaluative score. With a focus on nuclear power plants that use water-based cooling and heat transfer methods, the tool considers environmental factors of the site and how they relate to nuclear power plant supportability; it reviews the socioeconomic factors of the region surrounding the site; and it considers the factors of the location that can impact the operational technical efficiency of a nuclear power plant. Using both spectral and binary scales, where a spectrum is assigned a minimum, maximum, and most preferred value, and the binary inputs allow for yes or no responses, as well as a weighting scale for each individual input factor and high level evaluation category, the tool provides a final score indicating how well the location should be capable of supporting a nuclear power plant.
Study on Emergency Action Levels of the Modular High Temperature Gas-Cooled Reactor
Authors: Zang Xiaochuan , Liu Tao
Published: October 2017

The emergency action level (EAL) scheme for a modular high temperature gas-cooled reactor (HTR) plant refers to the generic EAL development guidance for pressurized water reactors (PWR) with HTR modification due to its design issues. Based on reactor’s accidents analysis and consequence assessment, EAL scheme of HTR is established through the steps of category and classification. Four emergency classes are set for HTR consisting of U (Emergency Standby), A (Facilities Emergency), S (Site Area Emergency) and G (General Emergency). The Recognition Category of Initiating Condition (IC) and EAL contains A - Abnormal Rad Levels / Radiological Effluent, F - Fission Product Barrier, H - Hazards and Other Conditions Affecting Plant Safety, S - System Malfunction. The methodology for development of EALs for HTR on Fission Product Barrier and System Malfunction has some differences from PWR’s due to differences on operating mode, inherent safety features and system characteristics.
Shape Characteristics and Condensation Heat Transfer of Supersonic Steam Jet in Subcooled Water
Authors: Fangnian Wang , Huan Qin , Yaodong Chen , Ning Bai , Mian Xing , Zhaocan Meng , Feng Shen
Published: October 2017

The shape characteristics and direct contact condensation (DCC) heat transfer of submerged supersonic steam jet were investigated. One of the shape identification methods is the DCC shape regime map as a function of nozzle exit pressure, mass flux and pool water temperature, another one is the Look Up Table. Then based on the theoretical analysis, the new basic expressions of supersonic steam jet dimensionless penetration length, maximum expansion ratio, and heat transfer correlations were given, which were in terms of Ma number and Ja number. The correlations were nonlinearly fitted and validated well against the experimental data form open literatures. The discrepancies of penetration length and maximum expansion ratio between predicted and experimental values were within ±25% and ±12% respectively. Supersonic steam jet DCC heat transfer is feasible to be calculated via the identification of steam jet shape and selection of the corresponding shape and heat transfer correlations.
HEPA Filter Clogging and Volatile Material Release Under Solvent Fire Accident in Fuel Reprocessing Facility
Authors: Ono Takuya , Watanabe Koji , Tashiro Shinsuke , Amano Yuki , Abe Hitoshi
Published: October 2017

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures.

One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters.

We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.
Transport Mechanism of Methane Steam Reforming on Fixed Bed Catalyst Heated by High Temperature Helium for Hydrogen Production: A CFD Investigation
Authors: Wang Feng , Zhou Jing , Wen Qiang
Published: October 2017

Performance of methane steam reforming reactor heated by helium for hydrogen production has been studied by numerical method. Results show with the increasing of reactant gas inlet velocity, temperature in the reactor drops, leading to the decreasing of methane conversion and hydrogen production rate. Methane conversion, hydrogen production and hydrogen production rate rise with the increasing of reactant gas inlet temperature, while the increasing degree of system thermal efficiency reduces. Besides, with helium inlet velocity rising, temperature in the reactor increases and reaction in the reactor becomes more sufficient. Therefore, methane conversion and hydrogen production also increase when helium inlet temperature of rises, but its influence is weaker compared to that of helium inlet velocity. In the process of methane steam reforming heated by high temperature gas cooled reactor (HTGR) for hydrogen production, lower reactant gas inlet velocity, suitable inlet temperature, higher inlet velocity and higher HTGR outlet temperature of helium are preferable.
Development of a Sub-Channel Analysis Code “MIDAC” for Core Thermal-Hydraulic Design and Safety Analysis in PWR Plants
Authors: Yodo Tadakatsu , Takeda Naohiro , Iida Naoko , Kawachi Motoko
Published: October 2017

In PWR, a Departure from Nucleate Boiling (DNB) is one of criteria for the thermal-hydraulic design and safety analysis. A sub-channel analysis code calculates local coolant conditions to evaluate the PWR safety margins such as a DNB Ratio (DNBR). Mitsubishi Heavy Industries, LTD (MHI) has developed Mitsubishi Three Dimensional Drift flux Code for Analysis of Core Two-Phase Flow (MIDAC) that began the development since the 1990s which is a sub-channel analysis code for DNBR and Peak Cladding Temperature (PCT) evaluations. The code design is based on a drift flux model for the two-phase flow and a radial heat conduction model for the fuel rod temperatures. MIDAC has been verified by comparisons with exact solutions and other codes, and validated by comparisons with test data based on a Phenomena Identification and Ranking Table (PIRT) under the core thermal-hydraulic design and safety analysis conditions. As a result, MHI confirmed the applicability of MIDAC to PWR conditions in the thermal-hydraulic design and Non-LOCA.
Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition
Authors: Sakuma Wataru , Miyata Shinya , Maruyama Manabu , Ogawa Junto
Published: October 2017

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left.

In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants.

The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs.

In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.
The China-Foreign Comparative Study of Public Participation in Environmental Impact Assessment of Nuclear Power Project
Authors: Xinhua Lu , Jianshe Chai , Yuan Fang
Published: October 2017

Started with the concept and significance of Public Participation (PP), and based on the current situation of internet voluntary action of nuclear related affair in China, this paper indicated that in China, PP in Environmental Impact Assessment (EIA) of Nuclear Power Project (NPP) was inadequate and cannot play the appropriate role in decision-making of NPP. By surveying the situation of PP of nuclear power industry in the developed countries, which focused on the incepts in the legal basis in PP, the participants, the participation stages and the methods, some issues and flaws in PP of China were indicated. Meanwhile, some suggestions were proposed as a reference of PP in EIA of NPP in China and in favor of the development of nuclear power safety.
Analyses of the Off-Design Point Performance of a High Pressure Ratio Intercooled Brayton Helium Gas Turbine Cycle for Generation IV Nuclear Power Plants
Authors: Gad-Briggs Arnold , Pilidis Pericles
Published: October 2017

The Intercooled Cycle (IC) is considered as a viable alternative to the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR), based on recent studies in a Nuclear Power Plant (NPP) configuration, which showed plant efficiencies of above 45%. The main difference in configuration is it does not utilise a recuperator. For part load performance, it is expected that the components of the IC will not operate at optimum conditions as the characteristics change. Thus the ability to demonstrate viable part load performance becomes an important requirement for the IC. The main objective of this study is to derive Off-Design Points (ODPs) from a known Design Point (DP) for a temperature range of −35 to 50°C and COTs between 750 to 1000°C. The ODPs have been calculated using a modelling & performance simulation tool designed specifically for this study and aim to provide a set of points that give operational equilibrium, which is critical to the economics of the plant. Results show that the intercooler alters the actual mass flow rate and compressor pressure ratio but the delta across an analysed range of 1 to 5% pressure loss shows a change of ∼9% in plant cycle efficiency, in comparison to the ICR (6%). Furthermore, the reactor pressure losses for IC has the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic trend maps have also been produced for the intercooler operation and the reactor and are applicable for NPP first order calculations. To that effect, it is also proposed to consider the intercooler pressure loss as a handle for ODP performance calculations. The analyses intend to bring further attention to the IC an alternative to current cycle configurations and to aid the development of cycles for Generation IV Nuclear Power Plants specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.
The Radioactive Aerosol Monitoring of the First Hot Test at China Reprocessing and Radiochemistry Laboratory
Authors: Xiaoyan Jiao , Yang Bai , Haoran Ma , Xin Xu , Ning Liu , Xiaorong Wang , Weimin Li
Published: October 2017

The first hot test have been completed at China Reprocessing and Radiochemistry Laboratory in September to December 2015. According to the technology process and radiation protection requires, the radioactive aerosol monitoring and airborne radioactive effluence monitoring was taken in workplace. During the hot test, the maximum concentration of α and β radioactive aerosol in controlled area I was 9.66Bq/m3 and 1.33×103Bq/m3, and the concentration of α and β radioactive aerosol in supervised area and controlled area II was under the authorized limits, which was 1.0–6.0×10−3Bq/m3 and 1.0–7.0×10−2Bq/m3. The leakage of radioactive aerosol from sealing chamber was not found. The concentration of airborne radioactive effluence in supervised area reached the discharge standard, which was α: under 7.0×10−3Bq/m3 and β: under 3.0×10−2Bq/m3. For the dissolution process, the concentration of 85Kr and 129I was significant parameters to monitored.
Research on the Reliability of Digital Instrumentation and Control System of Nuclear Power Plant Based on Dynamic Flowgraph Methodology
Authors: Zhuo Chen , Bo Zhao , Jian Yang , Jin-long Sun
Published: October 2017

With the development of information and computer technology, the Digital Instrumentation and Control (I&C) System has been widely used in nuclear power plants, which leads the tendency of NPPS’ construction and rebuilding on digital I&C system. As an approximate approach, conventional fault tree approach has been used quite often in the analysis of nuclear power plants’ Probability Safety Assessment (PSA), which combine with system components’ failure modes in order to modeling the digital system’s failure. However, for the reason that conventional fault tree approach has a great disadvantage on analyzing the reliability of digital I&C system, which may not be able to fully describe the dynamic behavior of digital I&C system with significant hardware/software/human action process interaction, multi-failure modes and logic loops, it cannot carry on effective modeling and evaluation of digital I&C system. Therefore it is necessary to establish some dynamic approaches to modeling digital I&C system. As a new probability safety analysis method, Dynamic Flowgraph Methodology (DFM) can model the relationship between time sequence and system variables because of its dynamic property. Therefore, DFM can be used to analyze the impact of software failure, hardware failure and external environment, which are closely related to the reliability of the whole system. In the first place, this paper introduces the theoretical basis, model elements and the modeling procedures of DFM and demonstrates how Dynamic Flowgraph Methodology (DFM) can be applied to Reactor Protection System with interactions between hardware/software and physical properties of a controlled process. Meanwhile, in this case, DFM and fault tree methodologies are both used to conduct the PSA for the same top event by calculating the probability of it and finding out the prime implicants of DFM and minimal cutsets of conventional fault tree. During the process of analysis, we mainly evaluate the reliability of reactor trip function of Reactor Protection System (RPS) by using DFM and conventional fault tree approach and mainly focus on modeling the four-way-redundant voting logic and the reactor trip breaker logic. Finally, through the comparison of this two methods and model results, it is concluded that there is a distinct advantage of DFM over conventional fault tree approach by using multi-logic to fully display the fault mode and utilizing decision table to describe the interaction between software and hardware. In general, conclusion can be drawn that, as a dynamic approach, Dynamic Flowgraph Methodology could be more accuracy and effective than conventional fault tree approach in analysis, ensuring the reliability and safety of the whole digital I&C system.
Single Event Effect Characteristics Analysis of Typical Circuit Elements in Spacecraft Power Systems
Authors: Wen Zhao , Chaohui He , Wei Chen , Xiaoqiang Guo , Peitian Cong , Fengqi Zhang , Rongmei Chen
Published: October 2017

Single event effect occurs when a single energetic particle penetrates sensitive nodes and deposits enough charge by ionization in semiconductor devices. It has become a major reliability concern for spaceflight. Single event effect characteristics analysis methods based on simulation are presented for typical circuit elements in spacecraft power systems. The failure mechanism and impact factors of single event burnout in trench power MOSFETs are investigated through numerical simulation. The broadening, quenching and capture of single event transients in combinational logics are analyzed by circuit simulation based on a coupled single event transient injection method. A behavioral modeling is introduced to predict single event effect sensitivity of the parallel to serial conversion circuit. The results show that ion-induced holes can spread efficiently to turn on the parasitic bipolar junction transistor and result in stronger carrier multiplication when the ion strikes at the center of the gate region in trench power MOSFETs. Increasing the depth of P+ plugs, decreasing the doping concentration of source regions and using lower drain bias voltages can mitigate single event burnout susceptibility. Using the same load capacitance for each stage inverter and selecting a suitable gate-width ratio are recommended to prevent single event transient broadening in inverter chains. Moreover, the reset pin of D flip-flops in the parallel to serial conversion circuit should be hardened by design according to the behavioral modeling results.
A New Effective Numerical Method for Leakage Rate Prediction of the Metal Gasket Sealing Assembly
Authors: Lin Fang , Guoliang Xu , Huang Xiaoming , Hao Huang , Yang Du , Yiping Zhu , Wenjian Yang
Published: October 2017

In order to evaluate the effect of roughness surface and its mechanical deformation on the leakage rate of gasket seal structure, the concept of effective separation height was introduced to characterize the characteristic scale of interface leakage channel. The effective separation height can be achieved by finite element analysis of the mechanical contact deformation of a single rough peak combined with the Roth contact model. The numerical results show that the leakage rate obtained by parabolic roughness is close to the experimental value, and the influence of the roughness angle on the settlement is very small. On the basis of the effective separation height, this paper realizes the theoretical prediction of the leakage rate by combining the rough surface computer generation technology and the numerical simulation of the leakage channel flow. The prediction results are in good agreement with the experimental results. The research of this paper has important guiding significance for the optimization design of the sealing performance of metal gasket.