Risk Assessment,Management and Application in Nuclear Power Plant Operation

To introduce the basic concepts of technical specification of nuclear power plant,a risk assessment and management technique based on the probabilistic safety analysis( PSA) method was proposed. The risk-informed method was used,and an example was given to show how to use some specific risk metrics like CDF / LERF /ICDP / ILERP to analyze and manage the risk associated with activities in nuclear power plant operation. The advantage of this technique can be concluded from this paper,and this technique should be used more widely and deeply in nuclear industry.

Checking of seismic and tsunami hazard for coastal NPP of Chinese continent after Fukushima nuclear accident

A checking on seismic and tsunami hazard for coastal nuclear power plant (NPP) of Chinese continent has been made after Japanese Fukushima nuclear accident caused by earthquake tsunami. The results of the checking are introduced briefly in this paper,including the evaluations of seismic and tsunami hazard in NPP siting period,checking results on seismic and tsunami hazard. Because Chinese coastal area belongs to the continental shelf and far from the boundary of plate collision,the tsunami hazard is not significant for coastal area of Chinese continent. However,the effect from tsunami still can't be excluded absolutely since calculated result of Manila trench tsunami source although the tsunami wave is lower than water level from storm surge. The research about earthquake tsunami will continue in future. The tsunami warning system and emergency program of NPP will be established based on principle of defense in depth in China.

Accident source term and radiological consequences of a small modular reactor

Considering the growing global demand for energy and the need for countries to achieve climate goals,there is an increasing global interest in small modular reactors(SMRs)and their applications.Accident source term and radiological consequence evaluations of SMRs are key components of nuclear and radiation safety reviews,which affect the site,exclusion area(EAB),and low population zone outer boundaries.Based on the design characteristics of the SMR and accident analysis results,a theoretical model of a whole-core fuel cladding damage accident was constructed to study the radioactivity released into the environment and its consequences.The accident source term and radiation dose calculation models were established to analyze the released amounts of radionuclides and the total effective dose affecting individuals at the site boundary.The results showed that the amount of radionuclides released into the environment after a whole-core fuel cladding damage accident reached 10^(14) Bq,among which the release amount of ^(133)Xe was the largest.The total effective dose at the site boundary 30 days after the accident was 8.65 mSv.The highest total effective dose affecting individuals occurred to the east-north-east.The results of the accident source term and radiological consequence provide technical support for site boundary dose assessments and reviews of SMRs.

Integrity-Management Characteristics and Efficiency Evaluation of Oil and Gas Pipelines

Pipeline integrity is a cornerstone of the operation of many industrial systems, and maintaining pipeline integrity is essential for preventing economic losses and ecological damage caused by oil and gas leaks. Based on integritymanagement data published by the US Pipeline and Hazardous Materials Safety Administration, this study applied the k-means clustering and data envelopment analysis(DEA) methods to both explore the characteristics of pipeline-integrity management and evaluate its efficiency. The k-means clustering algorithm was found to be scientifically valid for classifying pipeline companies as either low-, medium-, or high-difficulty companies according to their integrity-management requirements. Regardless of a pipeline company's classification, equipment failure was found to be the main cause of pipeline failure. In-line inspection corrosion and dent tools were the two most-used tools for pipeline inspection. Among the types of repair, 180-day condition repairs were a key concern for pipeline companies. The results of the DEA analysis indicate that only three out of 34 companies were deemed to be DEA-effective. To improve the effectiveness of pipeline integrity management, we propose targeted directions and scales of improvement for non-DEA-effective companies.

Response characteristics of PWR primary circuit under SBLOCAs considering steam bypass discharging

Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core safety.To explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power plant.Five different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was analyzed.The results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be guaranteed.The GCT was used in conjunction with the primary safety injection system during the placement process.When the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.

Radiation shielding design of a compact single-room proton therapy based on synchrotron

A synchrotron-based proton therapy(PT)facility that conforms with the requirement of future development trend in compact PT can be operated without an energy selection system.This article demonstrates a novel radiation shielding design for this purpose.Various FLUKAbased Monte Carlo simulations have been performed to validate its feasibility.In this design,two different shielding scenarios(3-m-thick concrete and 2-m-thick iron–concrete)are proved able to reduce the public annual dose to the limit of 0.1 mSv/year.The calculation result shows that the non-primary radiation from a PT system without an inner shielding wall complies with the IEC 60601-2-64 international standard,making a single room a reality.Moreover,the H/D value of this design decreases from 2.14 to 0.32 mSv/Gy when the distance ranges from 50 to 150 cm from the isocenter,which is consistent with the previous result from another study.By establishing a typical time schedule and procedures in a treatment day for a single room in the simulation,a non-urgent machine maintenance time of 10 min after treatment is recommended,and the residual radiation level in most areas can be reduced to 2.5 lSv/h.The annual dose for radiation therapists coming from the residual radiation is 1 mSv,which is 20%of the target design.In general,this shielding design ensures a low cost and compact facility compared with the cyclotron-based PT system.

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.

Standardization of Tritium Water by TDCR Method

The triple-to-double coincidence ratio （TDCR） method of liquid scintillation count- ing is an absolute measurement method of radioactivity. The formulation of the TDCR method and the established TDCR liquid scintillation counter are presented in this paper. The NIST standard reference material （SRM） of tritium water was measured to verify the performance of the TDCR liquid scintillation counter.

Comparison of two great Chile tsunamis in 1960 and 2010 using numerical simulation

Great differences in hazard and losses were shown from two tsunamis, both generated in Chile, one in 1960 and the other in 2010. Numerical simulation was applied to the tsunami analysis. The fault dislocation of the seafioor was assumed to equal to the initial tsunami wave field, which can be calculated by the formula of fault dislocation in the elastic isotropic half-space. The linear long wave theory was used as the tsunami hydrodynamic model, and the finite difference method and leap-frog scheme were selected for solving the equations. The accuracy of the simulated results was verified by the observed data in five tide gauges. By means of two scenario tsunamis, the analytical results show that the earthquake magnitude, bathymetry in rupture zone and rapid release of warning information in 2010 tsunami are the main explanations of the aforementioned great difference.

Application of Peps in Stress Analysis of Nuclear Piping

According to the nuclear safety regulations, this paper discusses the mechanical analysis method for piping system. Peps program has advantages of stress analysis and evaluation for nuclear piping. First, this paper introduces the Peps software, and discusses the process of stress analysis and evaluation for nuclear piping using the general finite element software;Secondly, taking nuclear class 2/3 piping system as an example, it uses Peps4.0 program to calculate the piping stress in variety of working conditions, such as weight, pressure, thermal expansion, earthquake, time-history force, and etc. Finally, the paper calculates the maximum stress and stress ratio according to the ASME.

Analysis on SH-CCT curves of HD15Ni1MnMoNbCu steel for nuclear power station

The microstructural evolution and Vickers hardness measurement in the welding heat-affected zone （HAZ） of HD15 Nil MnMoNbCu steel for nuclear power station were investigated by Gleeble-3180 thermal mechanical simulator, and the simulated HAZ continuous cooling transformation curves （SH-CCT） were measured simultaneously. With ts/5 inereasing from 3.75 s to 15 000 s, the product was obtained martensite, bainite, ferrite and pearlite, successively. The result of microstructure and Vickers hardness in the heat-affected zone was in good agreement with those measured by SH-CCT diagram with the heat input 16. 2 kJ/cm as an example to weld the HD15Ni1MnMoNbCu steel pipe using TIG/SMAW/SAW welding methods.

Stress Analysis for Reactor Coolant Pump Nozzle of Nuclear Reactor Pressure Vessel

Integrated reactor structural design makes the pressure vessel itself and loads more complicated, so stress concentration makes strength failure easier at reactor coolant pump nozzle. The general purpose finite element program ANSYS/ WORKBENCH was used for 3D stress and fatigue analysis and the results of the evaluation are based on RCC-M criteria. The integrated reactor structural design is evaluated to demonstrate with applicable criteria and ANSYS/WORK- BENCH has better operability than ANSYS APDL on stress analysis of reactor pressure vessel.

Dynamic scaling characteristics of single-phase natural circulation based on different strain transformations

To understand the dynamical system scaling(DSS)analysis theory,the applicability of DSSβ-andω-strain transformation methods for the scaling analysis of complex loops was explored.A simplified model consisting of two loops was established based on the primary and secondary sides of a nuclear reactor,andβ-andω-strain transformation methods were used to ana-lyze the single-phase natural circulation in the primary circuit.For comparison with the traditional method,simplified DSSβ-andω-strain methods were developed based on the standard scaling criterion.The strain parameters in these four methods were modified to form multiple groups of scaled-down cases.The transient process of the natural circulation was simulated using the Relap5 code,and the variation in the dynamic flow characteristics with the strain numbers was obtained using different scaling methods.The results show that both the simplified and standard DSS methods can simulate the dynamic characteristics of natural circulation in the primary circuit.The scaled-down cases in the simplified method exhibit the same geometric scaling and correspond to small core power ratios.By contrast,different scaled-down cases in the standard DSS method correspond to different geometric scaling criteria and require more power.The dynamic process of natural circula-tion can be simulated more accurately using the standard DSS method.

医疗垃圾焚烧飞灰处理研究进展

介绍了水泥固化、熔融、氯化焙烧及分步浮选等几种飞灰处理技术,着重分析了各项技术应用于医疗飞灰的可行性及局限性。水泥固化法因飞灰中高含量的活性炭和氯盐会阻碍水泥的水化作用,使水泥掺量过高,该方法不适合;熔融处理因飞灰量小、高氯、高碳会引起耐火材料的侵蚀及石墨电极的烧损等不利影响,不适合采用电弧炉熔融处理。医疗垃圾焚烧飞灰适宜采用残余炭式熔融炉;鉴于医疗垃圾焚烧飞灰中氯含量较高,直接焙烧处理是分离重金属的有效方法;分步浮选法是针对飞灰中碳组分含量高且二英在碳组分中富集的特性,通过两步浮选分离医疗垃圾焚烧飞灰中二英和重金属等毒害物,同时可脱氯。

核电厂一回路松脱部件监测系统的改进建议

针对我国压水堆核电厂一回路松脱部件监测系统(LPMS)存在的问题,分析了我国核安全法规标准以及国际上的相关监管要求,根据该系统的运行事件分析和经验提出了下一步系统设计及运行的改进建议,可供该系统的设计、运行及核安全审查参考。

Amidoxime-based adsorbents prepared by cografting acrylic acid with acrylonitrile onto HDPE fiber for the recovery of uranium from seawater

An amidoxime-based polymeric adsorbent was prepared by pre-irradiation grafting of acrylonitrile and acrylic acid onto high-density polyethylene fibers using electron beams,followed by amidoximation.Quantitative recovery of uranium was investigated by flow-through experiment using simulated seawater and marine test in natural seawater.The maximum amount of uranium uptake was 2.51 mg/g-ads after 42 days of contact with simulated seawater and 0.13 mg/g-ads for 15 days of contact with natural seawater.A lower uranium uptake in marine test can be attributed to the short adsorption time and the contamination of marine microorganisms and iron.However,the high selectivity toward uranium against vanadium may be beneficial to harvest uranyl ion onto adsorbents and the economic feasibility for recovery of uranium from seawater.

Flow-accelerated corrosion behavior of 13Cr stainless steel in a wet gas environment containing CO_2

This work investigated the flow-accelerated corrosion （FAC） behavior of 13Cr in a wet CO2-containing environment at different flowing gas velocities mid impinging mlgles, with the natural-gas pipeline environment simulated by a self-assembled impingement jet sys- tem. Surface molphology determination, electrochemical measurements, mid hydromechaziics numerical analysis were cmlied out to study the FAC behavior. The results demonstrate that pitting corrosion was the primary mode of corrosion in 13Cr stainless steel. High-flow-rate gas destroyed the passive film mid decreased the pitting potential, resulting in more serious corrosion. The corrosion degree witk various im- pact mlgles showed the following order： 90~ 〉 60~ 〉 45~. The shear force and the electrolyte from the flowing gas were concluded to be the determinm^t factors of FAC, whereas the shear force was the main factor responsible for destroying the passive film.

The Role of Atmospheric Teleconnection in the Subtropical Thermal Forcing on the Equatorial Pacific

The equatorial response to subtropical Pacific forcing was studied in a coupled climate model.The forcings in the western,central and eastern subtropical Pacific all caused a significant response in the equatorial thermocline,with comparable magnitudes.This work highlights the key role of air-sea coupling in the subtropical impact on the equatorial thermocline,instead of only the role of the ＂oceanic tunnel＂.The suggested mechanism is that the cyclonic （anticyclonic） circulation in the atmosphere caused by the subtropical surface warming （cooling） can generate an anomalous upwelling （downwelling） in the interior region.At the same time,an anomalous downwelling （upwelling） occurs at the equatorward flank of the forcing,which produces anomalous thermocline warming （cooling）,propagating equatorward and resulting in warming （cooling） in the equatorial thermocline.This is an indirect process that is much faster than the ＂oceanic tunnel＂ mechanism in the subtropical impact on the equator.

Numerical Simulation of PRHR System Based on CFD

In this paper numerical simulation of PRHR HX and IRWST is demonstrated using FLUENT, and different numbers of C-type heat transfer tubes and coolant inlet temperature’s effects for the residual heat removal capacity of PRHR HX, IRWST thermal stratification and natural circulation have been researched. It’s found that at a constant flow area when heat transfer tubes’ number increased outlet temperature of PRHR HX is lower, the whole water temperature of IRWST is higher, thermal stratification and natural circulation are more oblivious. At a constant mass flow when inlet temperature of PRHR HX increased, inlet flow velocity increases and outlet temperature is higher. But on the other hand the cooling rate increases at the same time, the average temperature of IRWST is higher, the range of thermal stratification expands and the velocity of natural circulation increases.