The Decommissioning Sub-Committee (DsC) of the Standard Committee of the Atomic Energy Society of Japan has been developing standards and guidelines for preparatory tasks (PTs) for decommissioning. The PTs consists of five sub-tasks, these are Radiological Characterization (RC), Investigation for the Plant Characteristics (IPC), Planning of Decommissioning Activities (PDA), Safety Assessment for Decommissioning (SAD) and Cost Estimation for Decommissioning. DsC has decided to publish guidelines on the first three sub-tasks: RC, IPC and PDA, and undergoes to develop them. RC and IPC, these two are collectively called as Plant Characterization (PC), would be carried out in the earliest stage of PTs. Results from PC are so called as “facts of the plant (FoP)” and referred by other sub-tasks such as PDA and SAD. Therefore, only reliable information from them could make the decommissioning plan safe and efficient. Especially, the reliability is one of the most indispensable factors to apply Graded Approach appropriately for every facility and action. RC would provide distribution of radioactivity inventory in a Nuclear Power Plant (NPP) and time dependent change of the distribution. A guideline on RC would provide procedures to evaluate the inventory. IPC would provide information about the NPP, such as information about facilities layout, design drawings of SSCs, operation records and other data necessary for decommissioning. The guideline on IPC would present two kinds of procedure to be carried out independently. One is surveying design drawings or usage and maintenance records to provide information about all SSCs and another is surveying the operation history to provide information about the operation history of the NPP. FoP resulted from PC are stored in a database system with appropriate record format. Here, the format shall be that data could be easily referred to by other sub-tasks and that update with the newly obtained data could be carried out accurately. The guideline of PDA would present procedures of planning the activities such as decontamination, dismantling and waste treatment. The guideline also presents a procedure of scenario making, namely scheduling the activities, and a procedure of making maintenance program of SSCs. Only careful considerations of the data resulted from PC in the database could make PDA appropriate. In order to effectively decontaminate and dismantle SSCs and treat the wastes, operators of decommissioning need to know character of each SSC precisely. Validity of the activities planned in PDA shall be confirmed in SAD and be reviewed as necessary.

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The Moody and Henry-Fauske critical flow models implemented in APROS have been validated against Marviken critical flow experiments and compared with other available simulations of the same experiments. Both models in combination with discharge coefficient 0.75 (suggested for best estimate calculations) produce results close to the experimental data for twophase flows, while one-phase flow of subcooled water is underpredicted. Using discharge coefficient 1.0 for subcooled water leads to a good match with the experimental results, while twophase flows become overpredicted.

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Sealed Radioactive Sources (SRS) are wildly used in industry, agriculture, medicine, education and research. Because of some reasons, for example the radioactive content decay, the new technology application or the operator bankrupt, disused sealed radioactive sources (DSRS) will be produced and shall be managed well because the activity of some SRS and DSRS is still higher and the potential risks may happened. SRS and DSRS must be sustainably managed for the protection of individuals, society and the environment. DSRS management options usually include transfer to SRS producers/original suppliers, long term storage, disposal, reuse and recycle, exempt or clearance. In Ghana, some DSRS have been transferred back to SRS producers/original suppliers, DSRS that cannot be transferred back to SRS producer/original suppliers are been stored in related facility. Ghana is planning to carry on DSRS disposal project with other countries. The disposal options include near surface disposal, borehole disposal, or geological disposal. An optimized disposal option in Ghana shall be verified considering the total amount of DSRS, the radionuclides of radioactive content, the safety and economy of the disposal options and other factors. Among these factors, safety assessment on DSRS disposal must be carried out. Usually, the software AMBER or ECOLEGO is used for the safety assessment on DSRS disposal This research work seeks to layout of safety assessment of disposal options of DSRS. DSRS options comparison.

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The integrated pressurized water reactor (IPWR) is a nonlinear, large-lag, and strongly coupled system with hundreds of measured parameters. The correlation between various parameters can be divided into strong correlation, weak correlation and irrelevance. Selecting the suitable number and type of strong correlation parameters to represent the problem is the main challenge to study system behavior prediction. Two different schemes for simultaneously performing the selecting feature and the training of predictive model of system behavior based on back propagation neural network (BPNN) are presented. The first scheme uses the covariance method (CM) to select parameters with large correlation coefficient as features. The second uses principal component analysis (PCA) method to reduce dimensions of raw data, the data with lower dimensions incorporates information of all measurement data. PCA has high selecting feature performances with low numbers of parameters. The measured data was obtained from the integrated pressurized water reactor RELAP5 program, then the data was standardized. After selecting features of two schemes respectively, the features are trained by the same prediction model. The effectiveness of the proposed selecting feature methods is tested on the integrated pressurized water reactor RELAP5 program. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of reactor core is predicted under load tracking conditions, the results show that as input data of the predictive model of system behavior, the data generated by the PCA method has lower dimensions, but has the same superior prediction performance.

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Hot Isostatic Pressing (HIPing) has been used by RollsRoyce to successfully manufacture nuclear pressure boundary components such as valves, piping, and pump casings. The majority of these components have been manufactured from stainless steels, typically 316L. There is also considered to be potential benefit from the HIP manufacture of other plant materials, including the large pressure vessel Low Alloy Steel (LAS) materials such as ASME SA-508. The advantages would include cost and lead time reductions, material quality, uniformity and inspectability. Applying the HIP process to LAS presents particular challenges compared to stainless steels. This is due to the propensity of LAS for oxygen pick-up, either in the powder manufacturing stage or subsequent can filling operations. The potential for oxide formation on the powder particles presents the risk of the material properties being adversely affected, particularly the fracture toughness which is extremely important in relation to the structural integrity of nuclear pressure vessels. This paper presents work conducted to assess the feasibility of achieving the required LAS material properties by HIPing to support nuclear pressure vessel manufacture. It presents tensile and Charpy impact toughness results, and makes comparisons with forged equivalent material. The LAS investigated was an ASME SA-508 Grade 4N model alloy. The paper shows that in all cases the HIPed material achieved higher tensile property values than the forged equivalent material. Although meeting the Rolls-Royce Charpy impact toughness target requirements, the HIP LAS Charpy Impact toughness values were lower than for forged equivalent material. From microstructural analyses conducted, it is considered that oxide decoration at the powder Prior Particle Boundaries (PPBs) is the main reason for the toughness difference to forged equivalent material. It is therefore proposed that in order to increase the Charpy impact toughness values to a level consistent with forged material, the oxygen level of the HIPed material will need to be reduced from the level achieved in this work, i.e. by either reducing the initial powder oxygen level or reducing further oxygen pick-up in the can filling operations, or both.

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Calculation capability of the pin-power distribution considering the variation of the assembly gap size due to assembly bowing was implemented in the pin-by-pin core calculation code SCOPE2. The previous studies show that the perturbation of geometry can be treated by the perturbation of macroscopic cross-section or atomic number density of the region instead of explicit consideration of geometry deformation. This methodology was applied to the assembly gap region: the variation of the gap size was treated by the correction on the macroscopic cross-section of the gap water. The correction model for the cross-section of gap water was implemented in SCOPE2 since it treats pin-by-pin cross-sections in a transport calculation. The correction was made according to the variation in the gap size. This implemented model has an advantage that the modification of the cross-section tables used in the core calculation is not necessary to consider the variation of gap size. For single assembly and multi-assemblies geometries, the assembly bowing model implemented in SCOPE2 was verified by comparing with the reference results using the assembly calculation code AEGIS, where the gap size perturbation was explicitly considered by varying the geometry of the gap region. It was confirmed that the variation of pin-power distribution due to the assembly bowing can be appropriately treated by SCOPE2 with the assembly bowing model.

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The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and Codes” is devoted to the development of tools for the evaluation of safety. WP-SO-2 “Experimental Programs and Operational Experience” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600, EBR-II and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as active and passive safety systems and other specific design features. This paper reports recent activities within the SO project.

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“Dead leg phenomena” is that the high-temperature fluid in primary loop heat the low-temperature static liquid in dead leg by convection heat transfer and heat conduction, and then the static fluid in dead leg is heated to evaporate or produce thermal stratification, which leads to the corrosion of pipe’s inner wall and valve. In this research, we use FloEFD which is a wellknown CFD software to simulate the thermal stratification in residual heat-removal system’s inlet pipeline and the heat exchanger’s outlet pipeline to get the fluid’s temperature distribution. The stream line in the pipe and the process of thermal stratification was investigated and the temperature distribution of the fluid was displayed. The calculation results show that the temperature in dead leg is lower than its saturation temperature and the design of this dead leg is appropriate. In addition, the calculation results prove that the thermal stratification exists in the dead leg. Finally, the formation mechanism of dead leg was speculated. The calculation results can provide a theoretical basis for the layout plan of residual heat-removal system’s inlet pipeline and heat exchanger’s outlet pipeline. The calculation and analysis results are of great importance for nuclear power plant operation safety and reliability.

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In a pressurized-water reactor(PWR) nuclear power plant, adjusting the boron concentration through Chemistry and Volume Control System(RCV) is an important assistant method to compensate for reactivity changes, and the coolant boron concentration has a very important impact on the design of Reactor Boron and Water Makeup System(RBM) and Boron Recycle System(ZBR). Therefore, the boric acid circulation study on RCV is very important for optimizing system and the equipment design for RBM and ZBR designers. In this paper, the transient work conditions and computing method on load following mode have been collated, and the accuracy of the calculation method has been tested on a PWR nuclear power plant in normal operation. A boric acid circulation analysis has been conducted on the new design of PWR units under construction, and the analysis results can provide a reference for the RBM and ZBR design. The main study results are as follows: The refueling cycle and burn up level have a great influence on the calculation results; The calculation method under load follow mode has been improved, and the accuracy of the results has increased by fifteen percent; The demineralized water tanks volume design largely depends on the boron concentration when the boron removal demineralizer begins to work; Due to the increase of core thermal power, the increase of the boron concentration in the primary loop and the lengthening of the reactor refueling period, the equipment volumes in the new design of power plants are larger than that of the operating plant. In order to improve the calculation accuracy and reduce manual burden, this paper uses Fortran to program and calculate.

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With the continuous development and progress of computer technology, nuclear power plant instrumentation control system will use more and more digital technology to replace analog technology. In order to ensure the safe operation of nuclear power plants, strict quality management is required for hardware and software in digital instrumentation control systems. Verification and Validation (hereinafter V&V) is an important means to ensure the quality of application software in digital instrumentation control systems. From the customer's perspective, choosing a suitable implementation team to complete the V&V work and effectively supervising it is a necessary condition to ensure the effective implementation of V&V activities and to ensure the quality of the products supplied. Based on the experience of previous nuclear power plant engineering practices and based on the study and understanding of laws and regulations and standards, this paper puts forward the common requirements that the author would insist on how to choose an appropriate V&V implementing organization to carry out the corresponding activities according to the regulations and standards, as well as the criteria for its judgment and the customer's supervision and management methods. It is of great value and significance for customer or general contractor of Digital Instrument Control System (hereinafter DCS) to select the suitable V&V implementation organization / team as well as the customer’s supervision and management.

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This paper first briefly describes the US maintenance rules, and the “Technical Policy for the Improvement of Maintenance Effectiveness of Nuclear Power Plants (Trial)” [5] issued by the National Nuclear Safety Administration (NNSA) in August 2017. The reasons to implement the Maintenance Rules (MR) at Qinshan NPP1 are also introduced. The important role played by PSA during the implementation of the MR project at Qinshan NPP1 was described in detail: 1) Determining the risk-significant SSCs; 2) Establishing SSCs performance criteria for reliability and availability; 3) Risk assessment and management of maintenance activities; 4) Periodically evaluating the impact of MR implementation on CDF.

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Traditional periodic maintenance techniques being employed in nuclear power plants usually fail to detect the potential degradation in performance of a sensor timely, and may increase workload and radiation exposure of the maintenance staff. The Redundancy Sensors Estimation Technique (RSET) to be presented in this paper is a noninvasive and in-situ monitoring technique based on measurement theory. The technique allows staff to monitor redundant sensors on-line and to assess their performance instantly. If such a REST can be applied to a nuclear power plant, it is expected that it will improve the safety of the plant and reduce the costs of operation and maintenance. This paper is concerned with a feasibility study on the REST’s application to a nuclear power plant. The results show that the RSET, featuring a strong generalization ability, can offer state estimation and fault diagnosis to signals from redundant sensors accurately. By adding drift data to the test dataset, the RSET can determine the signal drift accurately. The core algorithm of RSET can be explained by mathematical formulas and has high prediction accuracy. In conclusion, RSET can detect the performance degradation of redundant sensors in advance during the operation of plant.

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Advanced High Temperature Reactor (AHTR) is a type of the innovative Molten Salt Reactor (MSR) types which shows several promising characteristics such as high efficiency of thermal-electric conversion, high discharge burn-up and atmospheric pressure operation. However, due to the utilization of fuel assemblies with single enrichment, the power peaking factor (PPF) reaches to 2.09 with a radial PPF is more than 1.56, the high PPF is detrimental to the safety and economy of the reactor. We used a home-made AHTR-GATS code to minimize the PPF by radially varying the enrichment of fuel assemblies while keeping the FA average U-235 enrichment equal to the pre-design. To prevent the early convergence and obtain a better response, a novel hybrid algorithm which combines genetic algorithm (GA) with tabu search (TS) was developed to search for the best configuration corresponding to the desired patterns. The results indicated that compared with the pre-design the effective multiplication factor, k-inf of the optimized reactors were slightly improved and corresponding PPF were greatly reduced, and the temperature coefficient of reactivity were still negative throughout its lifespan. Moreover, the performance evaluation results of HAGATS show that it has strong robustness, high search efficiency and is suitable for loading pattern optimization.

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Japan Atomic Energy Agency (JAEA) has been conducting research and development for hydrogen production utilizing heat from High Temperature Gas-cooled Reactor (HTGR) which is expected to contribute in reduction of carbon dioxide emissions in various sectors including transportation, industrial, etc. Towards the realization of the nuclear hydrogen production, a safety design of coupling a hydrogen production plant to a nuclear reactor for a HTGR hydrogen production system should be established. One important consideration for the design is toxic gas leakages from the hydrogen production plant. The gas would spread to atmosphere by a leakage or a failure in the piping and equipment of hydrogen production plant. The released gas may flow into the control room in the reactor building through ventilation systems and affect its habitability. A safe distance between the reactor building and hydrogen production plant is required to prevent undue concentration increase of toxic gas in the control room. Traditionally, deterministic approaches are used to evaluate the adequacy of the distance with a worst leakage scenario for control room habitability, however, such approach may result in unreasonable plant design. With the aim of establishing a probabilistic approach for assessment of toxic gas leakage accidents in a hydrogen production plant to support risk-informed design of HTGR hydrogen production system, the present study focusses on development of an uncertainty analysis method for toxic gas concentration in a control room. The method consists of 6 steps; (1) Identification of uncertainty factors, (2) derivation of variable parameters, (3) identification of uncertainties in variable parameters, (4) identification of important factors considering the sensitivity analysis results and expert opinions, (5) uncertainty propagation analysis, (6) assessment of uncertainty analysis results. The method is then applied to representative toxic gas leakage accidents in a hydrogen production plant by thermochemical Iodine-Sulfur water splitting method coupled to the HTTR gas turbine hydrogen cogeneration test plant (HTTR-GT/H2 plant). The results obtained in the study leads us to the conclusion that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room.

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Japan Atomic Energy Agency has been developing an advanced aqueous reprocessing technology named “The New Extraction System for TRU Recovery” for spent Fast Breeder Reactor fuel. An effectively fuel dissolution should be required in the advanced process to introduce a crystallization technology which can separate about 70 % of uranium from dissolver liquor. We have focused Short Stroke Shearing technology of fuel pins bundle to promote the fuel dissolution in nitric acid solution. In this study, the feasibility of Short Stroke Shearing technology on simulated fuel pins for possible commercial reactor was discussed using the shearing device system on engineering scale. The length of sheared pins, released fuel ratio from the pin and shearing time are important as required performance in this technology, so the main shearing conditions such as shearing speed and magazine width were discussed to get the desirable performance. From the results of engineering test, it was found the higher pin occupied ratio in magazine is very effective for some required performances. It was demonstrated appropriate sheared conditions and the compatibility of the Short Stroke Shearing technology in advanced aqueous reprocessing method.

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Virginia Commonwealth University (VCU) is located in Richmond, Virginia, USA, and has one of the newest engineering colleges (founded in 1996) and nuclear engineering programs (founded in 2007). The nuclear engineering program is within the Department of Mechanical and Nuclear Engineering which has a large teaching and research portfolio with 521 undergraduate students, 106 graduate students, and 30 faculty. The department also has a nationally (United States) ranked online Master of Science degree program that has ~50 students enrolled in it. The graduation rates for nuclear engineering focused students at VCU are comparable to established United States nuclear engineering programs, including the University of California Berkeley, Purdue University, and the Massachusetts Institute of Technology. Nuclear engineering education at VCU goes beyond the classroom and research projects to include extracurricular activities. These activities include visits to nuclear facilities (e.g. Oak Ridge National Laboratory, Savannah River Site, and Vogtle Electric Generating Plant) and participation in professional societies (e.g. American Nuclear Society (ANS), Institute of Nuclear Materials Management, and American Society of Mechanical Engineers). Local chapters of these societies perform various public outreach events every year, including visits to K-12 schools. The ANS at VCU chapter has been especially active, hosting every year a “Girl Scouts Getting to Know Nuclear” workshop, “Boy Scout Nuclear Science Merit Badge” workshop, school teacher “The Science of Nuclear Energy and Radiation” workshop, and a visit to the Virginia State Capital to meet with Virginia Delegates and Senators to discuss the important contributions of nuclear science and technology for economic development and clean energy production. Due to this exceptional outreach and other factors, the ANSVCU chapter has been awarded the honor to host the 2019 ANS Student Conference, where public outreach will be an important aspect of the numerous events. The public outreach of VCU’s nuclear engineering program goes beyond the United States and includes activities in Egypt and India to support and strengthen security education and culture. Although VCU has one of the newest nuclear engineering programs in the world, they are already comparable by many metrics to established programs at other universities in the United States.

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The marine nuclear power plant is a floating nuclear power plant which supply power for the offshore oil drill platform. It is designed based on the standard and experience of the traditional onshore nuclear power plant. The reactor of the marine nuclear power plant is a 100 MWt PWR type small modular reactor with 2 loops. Each of the loops contains a main pump, a main check valve, and a steam generator. The pressurizer is set on one of the loops. The engineering safety features in the plant are somewhat different from those in the traditional plants. The special residual heat removal system, the passive residual heat removal, the square steel containment and containment suppression system are designed in the plant. The prevention and mitigation measures for severe accidents are set up on a reasonable and feasible basis to actually eliminate the large release of radioactive products. A MELCOR model and corresponding input deck were developed for the reactor coolant system, the secondary system, the containment system, the engineering safety features. Based on the safety analysis experience in the traditional nuclear power plant, the SBLOCA in the cold leg with the break diameter of 2.5 cm is chose as the initial event of the severe accident in the work. The sequence and important parameters in the accident are analyzed. According to the simulation results, the core exposed and heat up with the coolant release, and finally melt since the emergency core cooling system failure. However, the reactor pressure vessel (RPV) maintained integrity with the mitigation measure, i.e., the external vessel core cooling. The containment also maintained integrity, which prevented the large release of radioactive products to the other cabins and the environment. The work is useful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the SBLOCA induced severe accidents during the design process of the marine nuclear power plant.

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The nuclear safety is the basic demand for the development of nuclear energy. The severe accident which is the design extended conditions that can lead to core melt in the nuclear power plant, is of significant influence to the NPP design. The risk of severe accident in marine nuclear power plant (MNPP) is higher than that in the onshore NPP for the limitation of marine condition and the limited layout space. Thus, the severe accident mitigation measures are much more important in the design. The mature severe accident research method for onshore NPP is referenced for the MNPP research in the work. The reactor of the marine nuclear power plant is a 100 MWt PWR with 2 loops. The special residual heat removal system, the passive residual heat removal, the square steel containment and containment suppression system are designed in the plant. Three mitigation measures for severe accidents are set up on a reasonable and feasible basis to actually eliminate the large release of radioactive products. The pressurizer (PZR) relief extension is the measure to decrease the primary system pressure, avoiding the high pressure core melt. The external vessel reactor cooling (EVRC) is the measure to achieve the molten material in vessel retention, maintaining the reactor vessel integrity. The Passive autocatalystic recombiners (PAR) is the measure to decrease the hydrogen concentration in the containment, avoiding the hydrogen explosion. The MELCOR model and corresponding input deck were developed for the reactor coolant system, the secondary system, the containment system, and the mitigation measures. The typical severe accidents of the MNPP were chosen to evaluate the effect of three mitigation measures. The severe accident sequences and important parameters for the accident with and without the mitigation measure are comparatively analyzed. The influence of different mitigation measure to the key parameters and accident progression are achieved. The results showed that all the three measures could take the scheduled function, and are significant in mitigating the accident. The results can be used to support the design of severe accident mitigation measures in the new MNPP, which is meaningful to the development of the MNPPs.

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The four methods are used to simulate component cooling water plate heat exchanger in Fuqing unit 1&2 nuclear power plant, they are criterion number heat exchange formula, HTRI (Heat Transfer Research Institute) heat transfer factor empirical formula, Matin heat transfer factor empirical formula and A.Muley heat transfer factor empirical formula. The minimum difference of heat exchange area is calculated by using criterion number heat exchange formula compard with the operation data in Fuqing unit 1&2 nuclear power plant . So this method is used to analyse component cooling water plate heat exchanger in HPR1000 nuclear power plant, the heat exchanger supplied by factory comply with the engineering request,such as both in normal operation condition and in LOCA condition, the heat exchanger should remove enough heat to sea water. The problem in nuclear power plant that component cooling water subcool caused by sea water temperature subcool is a general issue for low-temperature site in winter. It supplied three measures to settle the problem, using the above method, the treatment of this problem for HPR1000 is chosed, and the calculation results of heat exchanger is given, the issues of plant unsafety operation and unplanned outage caused by component cooling water subcool is resolved.

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In nuclear power plant, By-pass System-A (named TSA) provides an artificial load during condenser steam dump unavailability or failure and in order to reduce the magnitude of temperature and pressure transients on the NSSS (Nuclear Steam Supply System). TSA system has some safety functions such as protection against overpressure or excessive heat up of reactor coolant system and protection against excessive cooling and safety shutdown and fast cool down. In TSA system, the most important equipment is steam dump valve which is powered by compressed air and could control flow mass through regulating valve travel. In order to know more about flow characteristics of TSA system, it's very important to know the relationship between flow mass, inlet pressure of valve (secondary loop pressure) and valve travel. In the paper, a model including steam dump valve, silencer and pipes is built and a numerical method is used to research the relationship between flow mass, valve travel and inlet pressure. During research, we found that outlet pipe reached a supercritical state when fluid is flowing from valve to silencer and a phenomenon of blocking flow is appeared in the valve. In Consideration of critical flow and blocking flow, we use a numerical simulation method to get the relationship between flow mass and valve travel and valve inlet pressure. The numerical method in this paper can provide ideas and basis for solving similar problems in engineering. The research results have an important reference value on selecting regulating valves.

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The current paper presents the work carried out in OKB "GIDROPRESS" to justify the separation characteristics of WWER-TOI steam generator (SG) using the STAR-CCM+ CFD-code. In the WWER-TOI project, new layout solutions were applied in the reactor plant (RP) as part of which the steam removal system from the steam generator was changed. Namely, in contrast to the WWER-1000 and WWER-1200 RPs where the steam removal was organized through ten nozzles combined into a steam collector, in the WWER-TOI SG the steam removal was arranged through one nozzle located on the cold collector side. This change leads to the formation of a non-uniform velocity field in the separation volume, between the evaporation surface and the distribution perforated plate (DPP), and can lead to the excessive increase of steam humidity. To ensure the steam separation characteristics of a horizontal steam generator with one steam nozzle, it was proposed to create a non-uniform resistance on the way of steam motion from the evaporation surface into steam nozzle applying a non-uniform degree of the DPP perforation. Two computer models of the SG steam volume with different steam removal schemes (through one and ten nozzles) were developed, a set of studies on verification and validation was carried out and a set of calculations were performed. According to the results of the calculations, the necessity of introducing a non-uniform degree of the DPP perforation was justified. Further, to determine the non-uniform degree of DPP perforation, a set of optimization calculations of the SG steam volume with one steam removal nozzle was performed. The non-uniform degree of DPP perforation of the WWER-TOI SG was selected, which provide steam velocity distribution as close as possible to the SG with ten steam nozzles. To justify the chosen design, sensitivity analysis was also carried out according to the hole diameters tolerance and steam load profile.

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The concept of multi-generation reprocessing in spent fuel for ceramic fast reactor is proposed in this paper. This reprocessing can deal with the spent fuel effectively with the simple and economic high-temperature dry (HT-dry) reprocessing without fine separation. After removing partial fission products from spent fuel by the HT-dry reprocessing, the new fuel can be used as the initial fuel for the next generation for the ceramic reactors, in which the fuel can be burned for decades. Based on the ceramic coolant fast reactor, this paper studies the neutron physics performance for multi-generation removing fission products. The parameters of the study includes Keff, beam density, nuclide mass. The results show the greater the removal amount, the higher the peak of the Keff, and the longer the duration of the continuous combustion. This preliminarily can prove the feasibility of the concept of this fuel cycle and meet the need of the nuclear energy strategy.

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Artificial intelligent (AI), as a mature high technology, had little used in nuclear power plant (NPP) field that it has a wide application scope to improve the management of NPPs. This paper analyzes the method and feasibility of using AI on field of NPP emergency operating guide, i.e. making the emergency operating strategy after accident occurred by AI system. This paper adopts artificial neural network (ANN) to calculate the emergency operating strategy, in which the post-accident status parameters vector is the input of the model and the emergency operating strategy is the output. In order to satisfy the requirement of ANN, emergency operating strategy is modeling based on the opinion that the fundamental strategy of an accident is to control the parameters which might effect on the safety functions. Controlling of the safety related parameters can also be deeply cataloged into control methods and system/component control, and forming an emergency operating strategy tree in the final. This paper also argues aspects of input data format, judgment of safety shutdown, ANN application arithmetic optimization, etc. A simple ANN model has been constructed in this paper and performing a practice calculation. The result verified that using AI on making decision during emergency condition is feasible. It is indicated that, currently AI system can provide auxiliary information for post-accident occasion, and this paper also illustrates the further research aspects for engineering practice on the next phase.

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If liquid sodium leaks as a liquid jet from pipe in sodiumcooled fast reactors, sodium droplets are produced during liquid jet impingement on the structures. Due to large contact area with surrounding air, generation of secondary droplets may lead to violent fire combustion. In the present work, high speed camera was first used to observe the liquid jet before the impingement. The splash ratio was then measured experimentally. It was shown that a phenomenological model using the impact frequency and the impact Weber number as the important variables can predict the splashing rate well. Distribution of the secondary droplet size was also measured by image analysis. It was indicated that the Sauter mean diameter of the secondary droplets was fairly proportional to the size of primary droplets.

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ALISA means a joint European-Chinese project named “Access to Large Infrastructures for Severe Accidents”. All parties can submit proposals to each other. The submitted proposal as below: In case of a core meltdown accident, molten corium relocated into the lower head of pressure vessel will be ejected into the reactor pit, if the lower head fails. Especially when invessel pressure is higher than the containment pressure, the ejected corium will be finely fragmented. Moreover, if the reactor pit is flooded with water during the vessel failure, fuel coolant interaction (FCI) may occur. The knowledge of FCI is insufficient and consequently the prediction models have large uncertainty. AP1000 use In-Vessel Retention as severe accident mitigation method. Water is injected to the channel between the RPV and insulation in the cavity. There is a quite small space in the cavity of AP1000. This proposal is to see what will happen if IVR fails. DISCO is a test facility special for FCI study in KIT that can simulate the high pressure corium eject from the reactor pressure vessel to the containment, which leads to high ejection velocity and high entry velocity into the water pool when the cavity is flooded, which is very unique for FCI investigation. Since the DISCO facility simulates also the containment, the experiment can also lead to better understanding on effects of FCI on the containment pressure and its integrity. Such experiment can be useful for FCI code validation.

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Dust re-suspension inside the vacuum vessel is one of the safety issues of ITER (International Thermonuclear Experimental Reactor). Plasma interaction with the PFCs (Plasma Facing Components) leads to their erosion, generating dust. One of the accident scenarios leading to dust re-suspension is the Ingress of Coolant Event (ICE) where a leak of the coolant pipes inside the vacuum vessel leads to flash atomization of the cooling water. The steam flow from the leak is considered to be the main source of dust re-suspension. Therefore, experimentations about the two-phase flow generated by the flashing liquid jet due to water leakage is important to identify the main physical phenomena involved in the aerosol particles re-suspension at low pressure. Flash-boiling experiments were conducted under primary vacuum conditions to investigate the flow behavior and structure of a superheated water (20°C to 140°C) injection into vacuum (1 mbar to 10 mbar). Using high speed dual-frame backlighted shadowgraphy and high speed PIV, qualitative and quantitative information were obtained on the two-phase flow such as spray geometry, droplets and gas velocities for different superheats and injected water mass flow rates. A transparent nozzle was also designed to better understand the phase change before the injection upstream the leak of coolant.

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Common cause failures (CCF) which probably consumedly reduce system reliability are important and complicated in system reliability analysis. GO methodology is an effective method for system reliability analysis even for a system with CCFs. Rationalization proposal was made to improve the accuracies of the previous algorithm. Whereafter a new algorithm was proposed. All common cause component groups (CCCG) can be dealed with together by a new GO chart to obtain system fault probability (FP) with only CCFs, then the system FP can be obtained. The new algorithm was then tested by analyzing two repairable systems with CCFs. All results consistently show that the new algorithm is a simple and available method for system reliability analysis with CCFs especially for the system with multiple CCCGs.

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Lots of porous carbon materials, including boroncontaining carbon (BC) and isostatic pressure graphite (IG-110), are used in high-temperature gas cooled reactor (HTGR). To prevent corrosion caused by carbon-water reaction at high temperature, water absorbed in the materials needs to be removed before the reactor operation. It has become the consensus of the researchers that the pore morphology of the material has an important influence on the physical adsorption and desorption process of the gas. Therefore, it is significant to explore the migration mechanism of moisture in carbon materials by studying the pore characteristics and hydrophilic properties of the material surface. Scanning electron microscopy (SEM) and contact angles tests were used to study the surface pore characteristic and hydrophilicity of BC and IG-110. The results of SEM scanning show that a large number of irregular micropores and cracks were found on the surface of BC, while the structure of IG-110 was more compact, and the size and number of pores were obviously smaller. 3 μL and 5 μL water drops were used to contact angle tests, it was found that the average contact angle of the BC surface with water droplets is 118.04°, which is greater than that of IG-110, measured as 112°. Finally, the water vapor absorption experiment of these two materials were carried out to verify that the pore structure and hydrophilicity have an effect on water absorption, and the pore structure has a more decisive effect on moisture absorption than the hydrophilicity.

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The MAAP5 code and CFD method are adopted to estimate the melting behavior of core peripheral components for an Advanced China PWR, called ACP1000 after a large break accident to determine the amount of stainless steel relocating into the corium pool. The results suggest that larger portion of peripheral components melting in CFD method because of considering the radiation of fuel rods on shroud and barrel before core melting and ignoring the cooling of steam on the components in core. Effects of different configurations of corium pool on heat flux imposed on the vessel wall are also investigated in this work. The heat flux in light metal layer for the corium pool that obtained by MAAP5 code could be higher than that of CFD case due to its thinner light metal layer.

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The China initiative accelerator driven subcritical system (CiADS) has already begun to build in the city of HuiZhou in south China’s GuangDong province in the year 2018, which is expected to be finished in the year 2025. During the next 6 years, a lot of fundamental work should be seriously considered and prepared, including how to establish a better way for workers and students to understand the structure of this nuclear facility, and how to effectively train the workers and students in order to accumulate the required resources to operate the CiADS. One better way to finish such tasks is the on-the-job training in a fast spectrum reactor, which has some common physical features as CiADS. However, in most instances, this kind of experimental reactor does not exist, and it will consume much costs with little flexibility. Another important reason making this approach not appropriate is that most training courses will expose these less skilled workers or students to the radioactive environment. Considering these situations, Institute of Modern Physics, Chinese Academy of Sciences (IMPCAS) has developed an Immersive Virtual Reality Platform (IVRP) to help the workers in IMPCAS and the students in the University of Chinese Academy of Sciences (UCAS) to better understand the project of CiADS. Although this special training system cost less to produce, it includes a lot of free combination elements and physical data, which are benefit for teachers to define different courses for their students by means of the attractive features provided by the IVRP.

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The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP Reference Male and Reference Female. In 2018 the Working Group 6 (WG6) of European Radiation Dosimetry Group (EURADOS) organized an intercomparison study on the usage of the ICRP voxel reference phantoms together with radiation transport codes. Organ dose calculation tasks were proposed in occupational, environmental, and medical dosimetry. The TRIPOLI-4 Monte Carlo radiation transport code has been widely used in radiation shielding, criticality safety and reactor physics fields for supporting French nuclear energy research and industrial applications. To enhance the application fields of TRIPOLI-4, the 2018 EURADOS-WG6 benchmark tasks are being taken into account by using different features of TRIPOLI-4 code. In this work, the ICRP reference voxel phantoms were first adapted into TRIPOLI-4. More than 14 million voxels were represented in a lattice geometry including 140 organs-tissues and 52 tissue media. Diverse exposure scenarios were then investigated by using ⁶⁰Co and ²⁴¹Am gamma-ray sources, ¹⁶N beta source, and 10 keV neutron source. The TRIPOLI-4 standard nuclear data library was utilized on these neutron, photon, electron, and positron coupled transport calculations. Energy deposition estimators for electron, neutron, and photon coupled with mesh tally options were used to estimate the organ absorbed dose D_T and the effective dose E. TRIPOLI-4 calculation methods and preliminary results for 2018 EURADOS-WG6 benchmark tasks are reported here.

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Small modular reactor (SMR) is a type of advanced reactor depending on its inherent safety, good economic performance, and multipurpose. SMR always adopts the multi-modular layout, which uses several nuclear steam supply system (NSSS) units to supply steam to one steam header. A type of multi-modular SMR system with pressurized water reactor (PWR) and once-through steam generator (OTSG) are studied. Due to the operation demands of continuity and flexibility during load following, a proper operation scheme of the multi-modular SMR must be presented. At first, the simulation platform for the multi-modular SMR are built. Three operation schemes are simulated and analyzed. (1) Balanced Scheme, which means every module undertakes the same load change during the load following operation. (2) Unbalanced Scheme, which means every modular undertakes the different load change during the load following operation. This operation scheme try to let more modules operate at full power to ensure its economy and operation simplicity. (3) Constant Output Scheme, which is mainly used during refueling process. When one module is shut down, other module undertake its load to ensure the total load keeping constant. Based on the simulation and analysis of these operation schemes, the control system is discussed.

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A novel method for assessing the reliability of 37- element CANDU [Canada deuterium uranium (reactor)] fuel was developed. The conceptual approach follows the principle of “best estimate plus uncertainty” (BEPU) where reliability or probability of failure to meet a fuel performance criterion is adapted as a measure of safety with due consideration of uncertainties. In this study, fuel performance was predicted using the industry standard code, ELESTRESS, which models fuel behavior during steady-state operation [1]. The outputs of the code were construed against failure criteria derived from industry norms to determine the probability of failure. In total, 105 independent iterations were made for each of 12 radially divided zones of a 480-channel CANDU reactor modelled after the Darlington Nuclear Generating Station cores [2]. The Monte Carlo simulation method is applied to analyze this problem. Probability distributions of fuel element related variables were estimated from real data supplied by Cameco Fuel Manufacturing (CFM) [3]. The inputs for element rating were simulated from a pool of core-following data generated using a 3D diffusion code, the Reactor Fuelling Simulation Program (RFSP) [4]. The results of the simulations predict zero fuel failures for all zones. The prediction of the method was compared to a deterministic, “limit of envelope” (LOE) benchmark, where all input parameters are assumed to be at their limits simultaneously. The benchmark analysis produced an output that is significantly closer to performance limits than those obtained from the proposed Monte Carlo simulation method.

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This paper presents recent results on high-temperature control blade degradation at the very beginning phase of a severe accident in BWRs. The large-scale experiment has been performed in JAEA/CLADS laboratory using Large-scale Equipment for Investigation of Severe Accidents in Nuclear reactors (LEISAN). A prototypic sequence of Fukushima Dai-Ichi Unit 2 has been taken. It has been shown that due to specific conditions happened at Unit 2 the lack of available oxygen allowed more flexibility during metallic melt relocation and heterogeneous redistribution of melt components due to specially recreated temperature gradient. Phase composition of remained B₄C control blade claddings at different elevations, and phase composition of melt has been investigated by complementary methods and have shown significant difference in elevations together with stratification of metallic components with origination of high-melting-point layers due to redistribution of steel components and involvement of B and C. It allowed absorber blade residuals with B₄C inside being severely damaged by melting to survive at 1477 °C and protect B₄C from direct contact with steam.

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In this paper, the driving power conversion rate of a twophase natural circulation loop is analyzed numerically. In operation of the two-phase natural circulation loop, part of thermal energy is converted into driving power, that is, the gravity difference between steam and water in vertical loops. The power drives circulation flow in this loop and it is converted into kinetic energy of the fluid, which is partially dissipated eventually with flow resistance. In steady state, the conversion rate is equal to the dissipation rate. In this study, modeling and simulation of a general two-phase natural circulation loop are performed. The dissipation rate of kinetic energy with flow resistance is calculated, which is equal to the driving power conversion rate in value. Furthermore, the conversion rate from thermal energy to the power is obtained with the conversion rate calculated indirectly and heat transfer rate of the heat source. Simulation results indicate that the conversion rate can be as low as 0.01%. Based on this result, design improvement of the natural circulation loop is proposed. In this design, the thermoelectric material Bi₂Te₃ is attached outside part of the heat exchanger tubes. With the material, the temperature difference between fluid inside and outside the heat exchanger leads to voltage, and the heat exchanger with the material actually works as a thermoelectric generator (TEG). In this way, part of thermal energy is converted into electricity with an efficiency which is much greater than that of original two-phase natural circulation loop. The electricity converted is further used to enhance circulation flow of water through a pump and natural convection outside the heat exchanger with a group of agitators. With this design, the driving power conversion rate of the two-phase natural circulation loop is increased, the driving force is enhanced, circulation flow and heat transfer can be improved.

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With regard to the design of a natural circulation system, the steady-state behavior of which is an important evaluate indicator, because it can directly reflect the flow capacity and heat transfer capacity. Previous scholars put lots of theoretical and experimental works on the steady-state characteristic of high-pressure natural circulation, while they done little on the open natural circulation, especially on the flashing-driven open natural circulation. In fact, the steady-state characteristic of the flashing-driven open natural circulation is much more complex, due to the existence of flashing in the riser, which leads the uncertainty of two-phase region length and void fraction distribution in the riser. On the other hand, the unambiguous mathematical expression of relation between mass flowrate and heating rate is hard to acquire, by the integral of steady momentum equation, due to the circuit integral quantity of accelerated pressure drop is not equivalent to zero. In this paper, the mathematical expression between the flashing- driven natural circulation mass flow rate and heating rate was put forwards by the circuit integral of steady momentum equation. The results showed good agreements with corresponding experimental data. The analysis results can provide the technical supports for the design of open natural circulation system.

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Passive safety systems, which do not need external input to operate, are widely used to enhance the inherent safety. They are receiving increasing attention due to the remarkable advantages in the simplicity and reduction of human interactions. The main task of passive system reliability assessment lies in the estimation of failure probability taking various uncertainties into account. However, it is usually a challenging work based on the traditional Monte-Carlo method because repeatedly running thermal-hydraulics code (e.g., RELAP5) is required. Unfortunately, it is computationally impractical because a large number of simulations are needed. In addition, each simulation of such thermal-hydraulic code may take hours, the computational cost of which is prohibitively high. In order to address this problem, in the present work we propose an adaptive sampling method for Kriging metamodel. The proposed approach selects samples strategically based on the information from the previous iteration. The proposed method is tested with reference to one benchmark case and then applied to the reliability assessment of the passive residual heat removal system in an integral type PWR (IPWR200). Results indicated that this adaptive method decreases the calls of performance function to construct a Kriging model and improves the calculational efficiency to a great extent.

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Source term estimation is one of the key steps for nuclear emergency decision-making, and it is usually of the most significant uncertainty and difficulty than the other steps. As revealed by Fukushima nuclear accident, we are so difficult to know what had exactly happened to the reactors due to the widespread unavailability of instrumental indications and power supply. As a matter of course, this dilemma leads to the hard decision-making about timely actions of emergency response based on the unconfident understanding of the reactors’ status and the radioactive releases to the environment. However, we’d like to insist that we unfortunately are not aware of the effective utilization of some existing valuable information sources such as Probability Risk Assessment (PRA) during the nuclear emergency response decision making. Hence, in this paper, we propose a new approach to estimate the emergency source term based on PRA knowledge and insights, including logic models and quantitative information (reliability data, release category and so on). The approach intends to intelligently diagnose the status of reactors and to give out the potential source terms in a fast and real-time manner, so that emergency response decision making can be significantly improved.

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This paper presents the hydrogen risk analysis of a large Pressurized Water Reactor (PWR) undergoing a severe accident using the HYDRAGON code. The code has been developed by the Department of Engineering Physics, Tsinghua University, as a hydrogen risk analysis software specialized for containment building of nuclear reactor power plants. The chemical interaction between zirconium cladding of the fuel rods and overheated coolant may release hydrogen when severe accidents occur in PWRs. The hydrogen mixed with air existing in containment may jeopardize the integrity of containment building if the flammable mixture explodes. In order to mitigate the hydrogen risk, many nuclear power plants install the Passive Autocatalytic Recombiners (PARs) that allow hydrogen and oxygen react chemically at low temperature without any sights of flame. Recombiners remove hydrogen and thereby prevent explosions. This work simulated the gas diffusion behaviors during the accident in the containment of a large PWR. The simulations obtained spatial concentration distribution of various gas species including hydrogen, oxygen, steam and nitrogen and several key parameters that are required to complete hydrogen burning and explosion investigation. Sensitivity analyses of recombiners’ quantity and positions and hydrogen discharge intensity provides the reference for hydrogen mitigation measure design in the containment. The results demonstrate that the number of recombiners rather than positions has stronger influences on hydrogen mitigation. Since the installation position reflects less sensitivity, the recombiners could be arranged as higher as possible if necessary.

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The lower head molten pool model is an important model for the In Vessel Retention (IVR) effectiveness evaluation. It has the characteristics of complex relationship, many input parameters and large uncertainty. The traditional local sensitivity analysis method has the disadvantages of large amount of calculation and low efficiency when performing sensitivity analysis of complex models. Based on variance decomposition of the global sensitivity analysis method, the sensitivity analysis library(SALib) and IVR analysis code(CISER) developed by Nuclear Power Institute of China(NPIC) are adopt to analyze the sensitivity of the input parameters for four key parameters, such as the lower head wall heat flux ratio. In the paper, the sensitivity coefficient and influence tread of input parameters on key results are obtained, which provides a reference for the optimization of the lower head molten pool model and severe accident mitigation strategies.

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The emergency operating strategy of design extension conditions (DEC) has become an increasingly important hot topic in the nuclear industry of China. The medium break loss of coolant accident (MBLOCA) with medium head safety injection (MHSI) unavailable is a typical and important condition for emergency operating strategy analysis. As MBLOCA results loss of reactor coolant inventory, the reactor core would be uncover if the inventory can’t be compensated in time, and lead to unacceptable results of emergency operating strategy, such as reactor core damage. Therefore, the inventory complement strategy is a key part of MBLOCA with MHSI unavailable emergency operating strategy. And the LHSI is the primary option in the inventory recovery strategies for MBLOCA with MHSI unavailable accident. However, establishing the flow rate of LHSI requires operator to depressurize/cooldown the primary system in a sharp rate, which will induce a pressure shock to primary system. Meanwhile, the emergency operating strategy should consider the MHSI maintain time, and not initiate the primary system depressurizing action unless there is void in the head of pressure vessel. Above all, the judgment of whether to depressurize/cooldown the primary system is the key of the emergency operating strategy for MBLOCA with MHSI unavailable accident. This paper raises two kinds of emergency operating strategy to deal with this accident. One is inventory complement strategy which is based on the outlet temperature of reactor hot leg judgment, while another is based on vessel level judgment. In order to evaluate the effectiveness of these two emergency operating strategies, a best-estimate calculation program is used to simulate a series size of MBLOCA with MHSI unavailable accidents. According to the calculate results, when the primary system depressurize/cooldown rate is fixed, for some special break size, the strategy which is based on outlet temperature of hot leg is too late to mitigate the reactor core inventory. While the calculate results of the strategy which is based on vessel level show it’s a timely and reasonable inventory restore strategy for MBLOCA with MHSI unavailable accident emergency operating strategy.

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Generally, plane strain assumption, i.e., infinite-length unsupported tube assumption is applied in the specialty cladding creep collapse analysis code. Actually the unsupported cladding tube length is usually less than 20 mm, and this real condition will result in lower creep rate than the one with infinite-length unsupported. In order to make the prediction better consistent with reality, the finite tube-length correction on creep rate, which is calculated based on infinite-length unsupported tube assumption, should be done. Two methods are introduced to calculate the finite-length tube correction factor based on 3D fuel rod creep collapse code ABCREEP. One method is called stiffness method, and it is developed based on that the stiffness difference between finitelength tube and infinite-length tube results in the creep rate difference. Other method is called lifetime method, and it is developed based on that the tube length difference between finite-length tube and infinite-length tube results in the creep collapse difference. The calculation investigation shows that, for stiffness method, the correction factor is mainly dependent on the ratio of unsupported cladding tube length and cladding mid-wall diameter. And for lifetime method, the correction factor is dependent on cladding material and the ratio of unsupported cladding tube length and cladding mid-wall diameter. When unsupported cladding tube length and cladding mid-wall diameter are both same, the correction factor of stiffness method is greater than lifetime method’s, i.e., stiffness method is more conservative than lifetime method in cladding creep collapse analysis.

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The effective thermal conductivity represents the overall heat transfer characteristics of a packed bed of spheres and must be considered in the analysis and design of pebble bed High Temperature Gas-cooled Reactor (HTGR). In this study, an experimental facility has been developed to measure the effective thermal conductivity of a simple cubic packed pebble bed. The effects of some key parameters including bed temperature and pebble materials on the effective thermal conductivity of pebble beds are investigated. The wall effect on the effective thermal conductivity is also analyzed. Meanwhile, the experimental results are compared with the ZSK correlation and the ZBS correlation. It is found that the effective thermal conductivity decreases with the increasing bed temperature. Moreover, the effective thermal conductivity of a bed filled with graphite pebbles is about 4 times that of stainless steel pebbles. The experimental results show that there is a reduction in the magnitude of the effective thermal conductivity at the near-wall region. The experimental results fit well with the ZBS correlation at lower temperature. While at the higher temperature, there is an obvious discrepancy between the experimental results and the two correlations.

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Heat pipe cooling technology can get efficient heat transfer performance due to its passive cooling design and phase change technology. The separated heat pipe shows great application prospects on nuclear security and economy improvement as a result of its flexible arrangement and excellent heat transfer performance. This paper carried out an experimental investigation on the heat transfer characteristics of the separated heat pipe which is used for spent fuel pool cooling. The experimental investigations explored the impacts on the heat exchange capability of separated heat pipe imposed by the volume loading, the initial vacuity and the temperature of condensing section under different conditions. The results indicate that the maximum heat transfer power of the separated heat pipes can reach 10kW/m²; the heat transfer performance first increase then decrease with the increasing volume loading, and the performance reaches the optimal when the loading volume is 1200 ml; higher initial vacuity and lower temperature of the condensing section promotes the heat transfer performance.

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Aerosol as the main component of fission products behaves strong diffusion and migration properties, thus it's important to prevent aerosol leak into environment. Pool scrubbing is a potential method to removal aerosol under accident conditions of the nuclear power plants. According to its solubility, aerosol can be divided into soluble aerosol and insoluble aerosol. The relative humidity of aerosol laden gas will increase during it passing through the liquid pool, which cause the hygroscopic growth of soluble aerosol and further affect the deposition characteristics of aerosol particles. This article studies the effect of operating parameters (liquid height and gas flow rate) on the aerosol removal efficiency by pool scrubbing including insoluble aerosol and soluble aerosol. The results show that the removal efficiency of aerosol particles is influenced by the liquid level and gas flow significantly. The removal efficiency increases with the increase of liquid height. Within a certain range of gas flow rate, the aerosol removal efficiency decreases first and then increases, tend to stable finally with the increase of gas flow rate. Besides, the hygroscopic growth of soluble aerosol particles will cause the removal efficiency curve shift to left and the most penetrate particle size decrease.

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In nuclear power plant, the reactor annular downcomer acts as the upstream of the reactor core, and boric distribution inside the downcomer is focused in the ECC situation. The relevant study indicates that the inlet coolant does not flow along annular channel directly, and the current multi primary loops layout makes the fluid mixing of the downcomer far more complex. In this paper, boron transport and mixing phenomenon inside the reactor annular downcomer were analyzed by Laser Induced Fluorescence (LIF) method, a reduziertreduced RPV model was built with transparent material, the boric acid was equivalently simulated by rhodamine B. The single loop and double loops flow mixing tests were respectively performed, the qualitative and quantitative data at cross sections were acquired. For the single loop test, the injected tracer transported along the circumferential position and was constantly diluted by the ambient pure water; meanwhile, a portion of the injection entered the lower space and filled up the annular channel in the same manner. From the time evolution of concentration gradient distribution, the increasement of inlet velocity is favorable to the uniform mixing. For the double loops fluid mixing tests, the tracer solution and pure water were injected into the model simultaneously through two independent entrances. The location of the intersection boundary of the two fluids was affected by the entrance momentum ratio, and the tracer distribution fluctuated remarkably in the lower space of the downcomer.

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NPP general operation means the start-up and shutdown operation process of the unit, and is an important NPP normal operation transient. Due to the numerous systems and complex interactions of nuclear power plants, the general operation strategy design of nuclear power plants is difficult. The traditional document-based design methods have weakness such as incomplete information expression, ambiguity, in-traceable, and difficulty in "demand-design" and "demand-verify" V&V processes. Nowadays the Model Based System Engineering (MBSE) has become a popular methodology for complex design work. This paper used the MBSE method to carry out the NPP general operation strategy design case and the design V&V process. Results show that this analysis method is efficient and well-organized.

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Graphite dust can be generated in the pebble bed hightemperature gas cooled reactors (HTGRs) through abrasion, corrosion, gas-to-particle conversion, radiation damage, etc. Understanding the dynamic behaviors of graphite dust is essential for reactor maintenance and safety issues. Among different dynamic processes, coagulation plays an important role in changing particle number concentrations and particle diameters. In HTGRs, particles in the reactor core and the steam generator are under very large temperature gradient. The thermophoretic effects result in relative motions for particles of different sizes that may lead to additional coagulation. Previous studies have shown that thermophoretic coagulation cannot be ignored in such extreme conditions. In this study, we calculate the thermophoretic coagulation kernel for small particles in the HTGR conditions and compare it with the Brownian coagulation kernel. The results show that the local temperature gradient and particle diameters play an important role in determining the relative importance of thermophoretic coagulation and Brownian coagulation. Then we simulate the aerosol evolution considering coagulation alone. The volume squared sectional method is used and the initial size distribution is a log-normal distribution. The simulation results show that the total particle number concentration decreases and the geometric mean diameter increases much faster for combined thermophoretic and Brownian coagulation than for pure Brownian coagulation under large temperature gradient. In addition, the combined coagulation process is much faster with a larger initial geometric standard deviation due to the differential characteristic of thermophoretic coagulation. These results indicate that thermophoretic effects can locally dominate Brownian coagulation in the HTGR conditions and should be considered in the modeling programs for aerosol evolution.

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On-line monitoring technology is researched by taking Qin-Shan nuclear power plant as the research object in this paper. The reactor thermal-hydraulic is simulated by THEARE which is one-dimensional thermal-hydraulic model, the reactor physics is simulated by REMARK which is three dimension space time Neutron Kinetics model, REMARK and THEARE are coupled on SimExec simulation platform to accomplish the feedback calculation of reactor thermal hydraulic and reactor physics. Reactor core safety parameters are determined according to the basic principles of reactor core safety, a graphical man-machine interface is made by C# program for displaying these core safety parameters. At last, the core on-line monitoring code is verified by different operate conditions, and the on-line monitoring results showed that the error between the results calculated by coupled model with the reference values from simulator is within the allowable range and the trend of parameters is consistent with the actual condition, which can illustrate that the core on-line monitoring code can satisfy the engineering demand.

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