In-vessel retention through external reactor vessel cooling (IVR-ERVC) is one of the key severe accident mitigation strategies, which takes away the heat of melt through the reactor pressure vessel external cooling and prevents the leakage of the radioactive materials. Studies have shown that SiO₂ nanofluids can significantly improve the critical heat flux (CHF) of the heating surface, and applying nanofluids to external cooling of pressure vessels can reduce pressure vessel failure risk. Sodium hydroxide (NaOH) as an additive in the spray water would change the chemical environment of cooling fluid, so this work investigated the effects of NaOH and Na₃PO₄ solution at PH 9.7 on SiO₂ nanofluid saturated pool boiling CHF under different concentrations and different downward heating surface orientation angle. The experimental results show that under the same concentration, the CHF of SiO₂ nanofluid based on NaOH and Na₃PO₄ solutions are nearly the same. Compared with SiO₂/H₂O nanofluid, the reduction in CHF is 15% at 0.01vol%. Same as SiO₂/H₂O nanofluid, CHF increased with the orientation angle of heating surface in NaOH and Na₃PO₄ environment. What’s more, scanning electron microscope (SEM) images were taken on the nanoparticle deposition layer formed on heating surface. The roughness, thickness, and the pattern of the nanoparticle deposition layer was different under different chemical environment, which lead to a decrease in CHF.

Modern optimization algorithms play an important role in solving practical engineering optimization problems, and are able to obtain better approximate solutions with smaller computational effort. However, for some scenarios where the computation is time-consuming, the computational effort of modern optimization algorithms is still relatively substantial, making it difficult to meet the requirements of fast optimization. This paper notes the unique advantage of data-driven machine learning-based algorithms, which can quickly make fast and accurate predictions on unknown data. This provides new ideas for improving the computational speed of optimization algorithms. This paper proposes a hybrid data-driven optimization algorithm that uses machine learning to replace the "learning mechanism" in modern optimization algorithms, giving higher speed and accuracy to the optimization algorithm. A comparison is made with various optimization algorithms on commonly used benchmark functions, and a practical engineering validation of the HDDOA algorithm is carried out on the optimization problem of the input time of mitigation measures for the containment of the IRIS reactor under SBLOCA accident.

Severe earthquakes are regarded as Beyond Design Basis Events (BDBEs) in nuclear safety. BDBEs require resilience to mitigate consequences of failures, i.e., to prevent catastrophic failure modes. The main vessel of fast reactor (FRV) is a thin-walled large-diameter cylindrical structure, of which buckling is expected as the critical failure mode during earthquakes. However, previous research indicated that no immediate collapse occurs after buckling. Therefore, the objective of safety design of FRV under severe earthquakes is to achieve a stable post-buckling state.

In current study, a Finite Element Method (FEM) model is set up on FINAS/STAR based on the conceptual design of pool-type fast reactor proposed by Japan Atomic Energy Agency (JAEA). Static and dynamic analysis are carried out to study the buckling behavior and post-buckling stability of FRV under horizontal, vertical and coupled loading. Buckling patterns and dynamic response characteristics of FRV under horizontal and vertical excitation are observed and analyzed. Fatigue damage assessment is carried out.

It is found that global stability can be expected after buckling occurrence. It is achieved by phase delay arising from degradation of stiffness and reduction of natural frequency of FRV. In addition, the effect of coupled loading is scrutinized. Compared with a pure compression loading, a coupled loading contributes to reduced energy cumulative rate and fatigue damage cumulative rate after buckling.

As mitigation strategies from structural mechanics for Beyond Design Basis Events (BDBE), authors propose failure mitigation methods. After the Fukushima Nuclear Power plant accident, such many efforts were made to mitigate BDBE consequences as portable devices, additional backup facilities and accident managements. In the structural mechanics field, however, efforts were only strengthening to prevent failures for both DBE and BDBE. This approach will lead to limitless requirements for strength and expensive plants.

As the break though approach in structural mechanics for BDBE, authors propose failure mitigation methods by application of the fracture control concept. The fundamental idea is the control of failure order and modes. Preceding failures release loadings and mitigate consequent failures. When preceding failure modes have small effects for safety performance, such as small deformation and clack initiation, and consequent ones are catastrophic modes as collapse and break, fracture control improve safety and resilience.

Failure consequence mitigation can be realized by passive characteristics of structures without additional equipment and electric power. Therefore, this idea enables simple and reliable plants

To demonstrate this idea, failure mitigation methods were applied to next generation fast reactors, where high temperature and low-pressure conditions.

In the case of loss of heat removal accidents, high temperature conditions accelerate creep deformation of structures. When deformation will redistribute loadings and reduce stresses at important portions such as cooling liquid boundaries, consequence to creep rupture of boundaries can be mitigated.

When excessive earthquake, plastic deformation and buckling become dominant, because of thin structures from low pressure. Above failure modes reduce rigidity and natural frequency. When the natural frequency becomes lower than the input frequency, vibration energy is hardly transferred to structures and mitigate consequent failure of structures, such as collapse and break.

Severe accidents can result in excessive pressure and steam build-up that can compromise the integrity of containment in nuclear powerplants. Fission products can get released into the environment after the breach of containment. Iodine is a major harmful radioactive fission product that can cause thyroid cancer if it gets released into environment and it has several exposure pathways. So, removal of iodine is necessary to keep the environment and people safe. Filtered Containment Venting System (FCVS) is a passive nuclear safety system that was developed to control over-pressurization, removal of radioactive iodine, and restrict its on-site and off-site release. Out of all forms of iodine (cesium iodide, elemental iodine and organic iodide), methyl iodide is most difficult to handle and remove by wet scrubbers. A lab-scale setup of FCVS was developed at PIEAS to carry out in-depth research on removal of methyl iodide. In this setup, compressed air was used to simulate high pressure in severe accidents. A sparger with 1 mm diameter holes was used for removal of methyl iodide in the bubble column. 0.2% sodium thiosulphate and 0.5% sodium hydroxide were used as scrubbing solution. An additional additive named Trioctylmethylammonium chloride (Aliquat-336) was also added as a catalyst. Hydrodynamic parameters play an important role in the working of a scrubber. Effect of different parameters such as throat gas velocity, concentration and gas holdup was studied. Comparison of removal efficiencies was done with and without the addition of Aliquat-336. Better results were obtained with addition of Aliquat-336. Overall, retention efficiency of >90% was obtained.

For the safety operation and cold shutdown of nuclear power plant, the I&C equipment is important to grasp the state of nuclear power plant. On March 11, 2011, at the Fukushima Daiichi nuclear accident that progressed to severe accident with core damage, many of the I&C equipment lost their functions and it was difficult to grasp the state of nuclear power plant. It is very important of measuring and grasping the parameters of nuclear power plant (e.g., RPV water level, RPV pressure) for accident management, at severe accident such as the Fukushima Daiichi nuclear accident.

In this paper, we introduce the actions for the I&C equipment for severe accident management in Kashiwazaki-Kariwa nuclear power plant units 6 and 7 (KK-6/7), based on the lessons learned from the accident.

Large break LOCA will cause an increase in containment pressure, containment temperature and containment radiation level in nuclear power plants (NPPs). Containment spray is one of the most effective ways to mitigate the consequences of large break LOCA for these following facts, first, with the large space containment design, the containment spray can decrease the pressure peak and keeps containment integrity. Secondly, the containment spray can decrease the aerosol radiation level in containment, iodine in particular, and reduce the risks of radioactive release. Above all, the common strategy of containment spray in NPPs generally includes automatic actuation with high spray flowrate, in order to achieve good results in relevant accident conditions. Meanwhile, the strategy to shutdown containment spray should be considered as a result of these facts that, a weakened effect in decreasing radiation will occur and negative containment pressure may cause containment integrity damage in post-accident long term operation. For the above considerations, the emergency operating strategy of containment spray based on radiation level in large break LOCA condition and the relevant best estimate work are studied based on one NPP in this paper, in order to achieve reasonable results in containment spray operating strategy, which are able to optimize containment spray and reduce the bad consequences.

The lower head of the nuclear power plant reactor pressure vessel (RPV) can be subjected to severe thermal and pressure loads during the event of a core meltdown accident. As effective accident management strategy, IVR-ERVC (In-Vessel Retention of molten corium through External Reactor Vessel Cooling) strategy is introduced to reduce the possibility of a reactor containment failure by terminating the severe accident progress inside a reactor. In this case, the mechanical behavior of the reactor vessel lower head is of importance both in severe accident assessment and the assessment of accident mitigation strategies. This paper proposed material constitutive model, which is extended from constitutive model of Takahashi [1]. The proposed model can predict stress under large range of temperature and strain rate, which in turn can be used to predict material deformation under IVR-ERVC strategy.

The suppression measures of steam explosions are essential for the integrity of a containment vessel in light water reactors during severe accidents. Polyethylene glycol (PEG) can potentially suppress steam explosions when a molten metal falls into a PEG aqueous solution pool. We also reported that PEG is a reliable additive to suppress spontaneous steam explosions with and without external triggers. Water with a PEG becomes cloudy when the solution temperature exceeds a specific limit, known as the cloud point. The dissolved PEG starts precipitating beyond the cloud point temperature. Such precipitates stabilize a vapor film around a high-temperature molten metal and prevent the fine mixing of molten metal in the solution. To evaluate the suppressive effect of PEG solution and the controllability of steam explosion, a small-scale experiment was conducted in which a prototypic reactor metal was melted and released into a solution pool. Type-304 stainless steel and 30 wt% Zr mixed with type-304 stainless steel were used as test metals. The test metal was melted by induction heating in a crucible and kept at 1700°C. The molten metal jet was immersed in a solution pool of 20°C. Visual observation of the interaction between the molten metal and solution shows the effectiveness of steam explosion retardant and the required PEG solution concentration for a molecular weight of 4 million grams per mol.

The continuous measurement of control rod positions can be achieved by the capacitance rod position sensor (CRPS), which can be used in the control rod hydraulic drive system (CRHDS). CRPS exists fringe effect in actual use, which will cause measuring error in the inlet section of the sensor. Fringe effect experiments of CRPS are carried out in this paper, and finite element model is constructed by COMSOL Multiphysics. Numerical simulation results are validated by experimental results. The quantitative characterization of fringe effect of grounding type CRPS is conducted, and the fringe error is limited by the method of resetting the starting point of rod positions. Results show that the fringe error can be limited in ±0.2 mm. This paper can provide a guideline for the engineering design of CRPS.

Adoption of SiC-matrix fuel elements in future pin-in-block type HTGR designs will enhance oxidation resistance of the fuel element in the event of the air ingress accident, one of the most worrisome accidents in HTGRs. This would eliminate the need for the graphite sleeves used in the current pin-in-block type HTGR designs and enable high power density core designs with sleeveless and directly coolable fuel structure. Such a concept itself has been suggested in the past. However, feasibility for a core design with the SiC-matrix fuel elements has not evaluated yet. The present work is intended to demonstrate the feasibility for a new core design upgraded from an existing conceptual core design, called HTR50S, with 50 MW thermal power and reactor outlet temperature of 750 °C. The new core design uses SiC-matrix fuel elements and increases the reactor power density to 1.2 times higher than the original HTR50S design. The calculation results showed that the new core design satisfied these target values on the reactor shutdown margin, the temperature coefficient of reactivity, and the maximum fuel temperature during normal operation.

At the Japan Atomic Energy Agency (JAEA), a design support tool for advanced nuclear reactors is currently under development. This tool is called ARKADIA-Design, and it is expected to support the integrated design evaluation of reactors from the viewpoints of safety, economy, and sustainability as a carbon-free energy source by utilizing NEW analysis/evaluation technologies such as AI, and the accumulated knowledge of fast reactor development.

One task for development of the ARKADIA-Design is to build a system that automatically identifies optimized design parameters by which an objective function specified by core performance is minimized (or maximized). In the present study, we set up a single objective optimization example problem with multiple constraints for a homogeneous two-region core of sodium-cooled fast reactor (which will be one of the target core types of the ARKADIA-Design), and showed that the optimal solution of this example problem can be automatically obtained by the Bayesian optimization method, which is a candidate optimization algorithm for the system. In addition, we also demonstrated how the system would assist the core design procedure in future, by indirectly solving a three-variable optimization problem of the core design. From these results and demonstrations, we confirmed that the system to be developed has potential as a useful support tool for the designers, enabling them to obtain optimal core designs efficiently.

To achieve an innovative core design, an optimization process for a core design has been developed as a part of the design optimization support tool named ARKADIA-Design. The core design optimization process integrated the core design analysis of neutronics, thermal-hydraulics, and fuel integrity and plant dynamics analysis with the Bayesian optimization (BO) algorithm is being developed. The BO can reduce the total number of iterative calculations and enhance the efficiency of the optimization. To establish a basic framework of the optimization process, a representative problem following an actual core design procedure was defined. Here, core design parameters are optimized to show a high core performance with inherent safety by preventing core damage in an unprotected loss of flow event of sodium-cooled fast reactors. In this study, to confirm an applicability of the optimization process with the BO algorithm for the representative problem, a single-objective optimization problem was solved by performing the integrated analysis only between neutronics and plant dynamics as a first trial. In addition, an effectiveness of the optimization process was discussed by comparing with an ordinary core design process. An optimal solution in the representative problem was acquired by performing the integrated analysis with the constraint BO.

Recently, we have been proposing a sleeveless SiC matrix fuel compact design to increase the core power density and enhance the safety for High Temperature Gas-cooled Reactor (HTGR). For that application, the oxidation behavior of Reaction-Sintered Silicon Carbide (RS-SiC) under conditions of air ingress accident needs to be investigated. However, few study focused on RS-SiC oxidation, especially under conditions of interest for HTGR. The detailed RS-SiC oxidation kinetics and passive-to-active transition behavior remain unclear. In this work, aiming at feasibility of sleeveless SiC matrix fuel compact for HTGR application, thermogravimetry (TG) measurements were carried out in the conditions of 800 ~ 1400 ℃ and 1 ppm ~ 20 % oxygen concentration under atmospheric pressure, 30 min real-time monitoring. As a result, we have experimentally determined the passive-to-active transition that might be encountered in case of accidents. The present study successfully collected basic data on RS-SiC oxidation, thus, the material feasibility of sleeveless SiC matrix fuel compact for HTGR use was obtained. We partly concluded that the RS-SiC can be applied to HTGR fuel matrix material in the case of air ingress accident.

High Temperature Gas-cooled Reactors (HTGRs) adopting silicon carbide (SiC) instead of conventional graphite for fuel compacts with TRISO fuel particles have been proposed to improve the power density and to enhance the oxidation resistance of the fuel elements in the air ingress accident. The thermal conductivity of the matrix for the fuel compacts is considered to be one of the critical properties for the nuclear design of HTGRs, and higher thermal conductivity of the fuel compacts is strongly desired. To apply SiC as the matrix for fuel compacts, SiC should be dense to retain the fission products inside of the fuel compacts and to increase the thermal conductivity. In this study, therefore, we selected reaction sintering method, which is capable of fabricating dense SiC at lower temperature in a short time, and the fabrication process of the dense SiC matrix fuel compacts by reaction sintering was studied. Furthermore characterization and thermal conductivity measurement of the reaction-sintered SiC were conducted. The reaction-sintered SiC ceramics were so dense and achieved the thermal conductivity of 64 W/m•K at room temperature and 26 W/•K at 1000 oC. XRD, SEM and EDS results suggested that the reaction-sintered SiC mainly consisted of SiC with small amount of residual Si.

In order to achieve carbon neutrality (CN) and energy security for Japan, Toshiba Energy Systems & Solutions Corporation (Toshiba) aims to supply an improved ABWR that can be constructed and operated in early 2030s.

In particular, the installation of the Specific Major Accident Response Facilities (SARF) is now mandatory for existing reactors as a countermeasure against APC (Airplane Crash), and the situation is very severe in which plant safety and economic competitiveness conflict with each other. Considering the situation, Toshiba has developed design and evaluation technologies for power plant structures that can achieve a high degree of compatibility between safety and economy.

The ability to precisely simulate the process from impact to structural failure will help increase confidence in the effectiveness of APC protective structures and improve the accountability of permitting.

Therefore, the analysis method was verified through experimental research and analytical research assuming APC, and by applying the verified method to structures such as the reactor building, it was confirmed APC protection is functioning effectively.

The design of robust reactor buildings based on this technology will prevent the occurrence of severe accidents caused by APC, thereby improving safety. The design of a robust reactor building using this technology will improve safety by preventing the occurrence of severe accidents caused by APC, and will also improve economic efficiency by eliminating the need for specific facilities to deal with severe accidents.

Westinghouse Electric Company LLC has an ongoing initiative to develop a fully in-house suite of software applications to manage nuclear plant design data. Westinghouse began this initiative due to the challenges of having to extensively customize commercially available data management software to suit specific business and process needs. In pursuit of a better approach, Westinghouse has developed a flexible application framework. This has enabled the building of discipline-specific applications that integrate with each other and external software tools, such as 3D modeling software. In the process, Westinghouse has faced several challenges, some of which are specific to the nuclear industry, such as meeting industry quality assurance requirements to maintain lifetime records in accepted file formats. Software functionality has been developed to address those requirements.

In recent years, the ratio of renewable energy has increased toward carbon neutrality, and the ratio of thermal power generation is expected to decrease in the future. Until now, nuclear power plants in Japan have been utilized as base-load operations only, but there is an increasing need to implement Flexible Power Operation (FPO) in nuclear power plants as an electric power grid stabilization measure against the introduction of renewable energy. In response to this situation, the new type of Pressurized Water Reactor (PWR) SRZ-1200 is employing a new control scheme to improve the performance of the FPO. The main development is the primary coolant average temperature control scheme with selecting the control bank to be driven so that the axial power distribution (ΔI) in the core is kept within an acceptable range by the control rod control system. By implementing this control scheme, it is expected to improve the FPO performance of the SRZ-1200 compared to the existing reactors in Japan.

Nuclear power generation is a practical option of clean energy that emits a few amount of CO2, and the Kansai Electric Power Group's Zero Carbon Vision 2050 includes promoting efforts toward improving the plant operation of existing NPPs and planning construction and replacement with advanced light water reactors. However, from the 2030s onwards, the number of reactors that reaches the statutory upper limit of 60 years will gradually increase, so that the construction of new reactors will naturally be necessary to cope with the reduction in nuclear power generation capacity. The construction of advanced light water reactors are project that takes more than 15 years of time with a large investment. In Japan, which experienced the Fukushima Daiichi accident, the concept and design of new reactors must be acceptable to all stakeholders for a long period of time. Safety measures were implemented for the reactors that were restarted with the goal of achieving the world's highest safety level. However, the concept of those safety measures were strongly influenced by the structure and layout of existing equipment. Therefore, they are not necessarily rational and optimized in consideration of operability and economy. For advanced light water reactors to be constructed, flexible measures can be considered from the design stage to meet possible needs. Therefore, we are proceeding with a rational and optimized design, and clarifying safety, economy, and operability requirements expected for practical use.

MHI has been developing an advanced light water reactor (LWR) “SRZ-1200”[1]. For continuous use of Plutonium (Pu), it is desirable for light water reactors to be capable of loading a large fraction of MOX fuel.

Conventional MOX fuel assembly for Japanese Pressurized Water Reactors (PWRs) is designed with all MOX fuel rods – unlike the ones for BWRs – with distribution of Pu content to mitigate the power peak in the assembly. Today, the MOX fuel assemblies have been loaded to several Japanese PWRs; however, a fraction of MOX fuel assembly loading is limited to the one fourth of the core (1/4MOX core) to ensure a sufficient shutdown margin. With the aim of increasing a fraction of MOX fuel in the core, it is necessary to enhance the control rod wort, which is a key parameter for shutdown margin, against MOX fuel assembly.

As one of the measures to enhance the control rod worth, the advanced MOX fuel assembly, which consists of both uranium dioxide (UO2) fuel rods and MOX fuel rods, has been developed in this study. Inside the advanced MOX fuel assembly, UO2 fuel rods are placed around the control rod guide thimbles, and a fraction of MOX fuel rods in the assembly is approximately 80%. In addition to the measure mentioned above, combination of Ag-In-Cd (AIC) and boron carbide (B4C) with enriched boron, which has larger neutron absorption cross section than AIC does, is used for neutron absorber material of control rod. By implementing both the advanced MOX fuel assembly and AIC-B4C combined control rod, this study confirmed that the core can be fully loaded with the advanced MOX fuel assemblies (i.e., over twice more of annual Pu consumption to the core compared to the 1/4MOX core can be achieved) while maintaining a sufficient shutdown margin.

Mitsubishi Heavy Industries Ltd. (MHI) has been developing a new pressurized water reactor (PWR) SRZ-1200 that will satisfy future social needs. The SRZ-1200 is a 1,200 MWe evolutionary medium-sized nuclear power plant significantly safer and more reliable than the conventional plants by taking into account the lessons learned from the accident at Fukushima Daiichi Nuclear Power Station (the Fukushima Daiichi NPS accident), and at the same time adequately addressing the future energy situation in Japan such as expansion of renewable energies. This paper introduces the design concepts and the plant design features of the SRZ-1200.

In the framework of the Horizon 2020 project PuMMA (Plutonium Management for More Agility), a burner version of ALFRED (Advanced Lead Fast Reactor European Demonstrator) was designed to provide a reference case scenario of a plutonium burner Generation IV LFR (Lead Fast Reactor). The design was developed to increase the sustainability and performance of the reactor, using MOX (Mixed Oxide) fuel with high plutonium content, up to 40 wt.%. The core modifications were implemented aiming at maximizing the plutonium burning performance while preserving the robustness and safety of the original design and allowing complete integration with the same primary system arrangement. The new core’s neutronics were modeled with two different codes (ERANOS and SERPENT 2), whose results were applied as inputs for the thermal-hydraulic characterizations of a hot single channel in RELAP5. Finally, the neutronics and thermal-hydraulics outputs were employed as input to assess the ALFRED-Burner pin normal operating conditions with the fuel performance code TRANSURANUS. The analysis of the best-estimate reference scenario was followed by a sensitivity analysis to determine which fuel and cladding property parameters had the highest impact on pin performance results. The verifications performed so far confirm the promising safety features of the lead-cooled fast reactor and its performance with high plutonium content MOX fuel. However, the outcome of the analysis also highlights the potential for improvements in the fuel performance code. Such developments would allow enhanced performance and safety assessments.

Clarifying the dynamic buckling behavior of PWR steel reactor containment vessels (PWR-CVs) subject to seismic excitation is critical for improving a numerical seismic design scheme for PWR-CVs. In evaluating the static elastic-plastic buckling strength for PWR-CVs by a finite element (FE) analysis, first we compute the moments and axial forces acting onto the multiple lumped mass beam model of a targeted PWR-CV from the linear dynamic analysis, second using the buckling evaluation formula on JEAG 4601 addenda-1984 we calculate the equivalent static load distributions in terms of shear and axial forces as input loads to the FE model. This static FE buckling analysis for a PWR-CV idealizes dynamic buckling phenomena as be static phenomena, so we cannot basically consider the effects of dynamic behavior on the buckling strength of a PWR-CV. To tackle more realistic buckling assessment of a PWR-CV, we carried out the linear dynamic analysis for a PWRCV FE numerical model and discussed the features of the response considering the oval vibration modes. It was revealed that the response accompanying the oval vibration modes showed the lower value compared with that from the static FE buckling analysis. In contrast, in a previous study by Hidaka et al., 2022, it was discussed that the buckling load of static FE buckling analysis for a PWR-CV is higher than the buckling load of buckling evaluation formula on JEAG4601. As a result, the ratio of buckling load to response in the linear dynamic FE analysis was within 3% of that in the static FE buckling analysis.

A parametric study has been an effective tool in thermodynamic modeling and simulation for investigating the effect of individual variables on a complicated model. In this study, the parametric study is conducted on a novel Organic Regenerative Transcritical Cycle (ORTC) that is integrated with a Small Modular Reactor (SMR). The eight independent variables include: low-, mid-, and high-pressure mass ratios, low-, mid-, and high-pressure ratios, maximum temperature, and pressure, as well as the number of discretized sections in the primary heat exchanger. In order to focus on viable designs, the design space for each independent variable is limited to no and near-zero penalty domains. The results provide a clear trend on how each independent variable impacts the system’s net thermal energy into the power cycle, net power out, and the first and second-law efficiencies. In addition, the temperature profile in the primary heat exchanger for the current Regenerative Rankine Cycle is compared with that of the proposed ORTC. The effects of single and double-sized primary heat exchanger for both cycles are also compared. As the result, the proposed ORTC can potentially outperform the current Regenerative Rankine Cycle under the design conditions and provides an alternative option for small modular light-water reactors.

The new generation of nuclear energy systems represented by liquid metal reactors has a complex process of multi-physics mutual coupling in the process of steady-state or transient operation. In addition to the neutron feedback of the coolant in the core, there are some complex flow heat transfer phenomena in the reactor. This will affect the thermal stress distribution of the structural components in the reactor, and then change the size and physical properties of the core flow channel. Therefore, it is often necessary to couple physics, thermal engineering, and structural mechanics in the design of new reactors. To quickly and accurately calculate and predict the operating state of nuclear reactors online, this paper uses machine learning combined with multi-physics coupling prediction for the digital twin of reactor operation to achieve high-speed precaution on neutron flux distribution, power distribution, and temperature fields. Then we build a multi-physics fast computing model based on model reduction technology and machine learning to achieve physical guidance. The predicted physical fields are proven to achieve high accuracy in a short time. This paper is of great significance for the design, development, operation, and subsequent realization of digital twin technology of new reactors.

Using S-CO2 to replace the current core cooling medium of helium cooled reactor and the traditional steam power conversion working medium, and gas foil bearing to replace the traditional bearing can make the compressor, gas turbine and other power system equipment compact structure and small volume, reduce the construction cost of nuclear power plant, realize the modular construction technology, shorten the construction period of nuclear power plant. Based on this background, the gas foil bearing-rotor system was studied, the centralized load method and the piecewise modeling method were used to model the rotor subsystem, and the influence law of shafting motion parameters on rotor dynamic behavior was obtained. The research results can provide an important reference for the subsequent rotor system design.

Japan Atomic Energy Agency is developing an evaluation system aided by artificial intelligence (AI) named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle), aiming to offer solution for challenges to the design and operation of a nuclear plant. A sub-system of it, named ARKADIA-Design, is being developed to support the design optimization study for an advanced nuclear plant including a sodium-cooled fast reactor (SFR). Authors are developing a design optimization process for the structure of the component in SFR. This paper describes the outline of a design optimization process, the brief introduction of evaluation methods for the process, and the result of the demonstration of the optimization process for a feasibility study. The development is being performed in a representative problem considering the thermal transient and seismic motion as a major issue in SFRs. In the optimization process, the failure probability is used as the element of the objective function for optimization, to compare the contribution rate of the loads to the failure that occurred in different mechanism in an equation. Since a number of evaluations for the uncertainty quantification in the estimation of the failure probability is needed, the simplified evaluation method for failure probability in the optimization process is being developed to reduce evaluation cost. And iterative evaluations for updating variables of design parameters to achieve the optimal solution are needed. The automatization function in the optimization process is being developed for efficient execution of the optimization process. Through the demonstration, it was confirmed that the optimization process under development may provide an optimal solution to the representative problem.

The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake by applying the failure mitigation technology. This study regarded those measures for improving resilience of important structures, systems, and components for safety to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Accident sequences leading to loss of decay heat removal are significant contributor to seismic CDF of sodium-cooled fast reactors (SFRs), and those sequences result in core damage via ultra-high temperature condition. This study improved the methodology to evaluate not only the measures against shaking due to excessive earthquake but also the measures at the ultra-high temperature condition. To examine applicability of the improved methodology, a trial calculation was implemented with some assumptions for a loop-type SFR. Within the assumption, the measures for improving resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. Through the applicability examination, the methodology for the effectiveness evaluation was developed successfully.

Quantitative assessment of the effect of accident management on the various external hazards is essential in the nuclear safety analysis. This study aims to establish the dynamic probabilistic risk assessment methodology for sodium-cooled fast reactors that can consider the transient plant status under continuous external hazards with corresponding countermeasures operating stochastically.

Specifically, the Continuous Markov chain Monte Carlo (CMMC) and Deterministic and Stochastic Petri Nets (DSPN) methods are newly applied to the severe accident analysis code, SPECTRA, which can conduct dynamic plant evaluation in the different severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. In the DSPN-CMMC-SPECTRA coupled frame, the latest safety functions of the plant components/systems can be stochastically determined by the DSPN-CMMC grounded on the current plant states under continuous hazard and the interaction between the multi-state components/systems; then, SPECTRA can evaluate the following plant state determined by the latest safety function of the components/systems. Therefore, the advantage of this newly developed DSPN-CMMC-SPECTRA frame is having the capability to quantitatively and stochastically evaluate the transient accident progressions that potentially lead to the core damage under the continuous external hazard scenario.

As for the preliminary exam on the DSPN-CMMC-SPECTRA frame, one of the typical external hazards of continuous volcanic ashfall is selected in this research. In addition, the numerical investigation of alternative accident management' effects has also been carried out and quantitatively confirmed in this research.

The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor “Joyo” in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, a level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD level 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As the result, the frequency of each accident sequence group decreased significantly, and total frequency of the accident sequence groups decreased to about 1×10-6 /reactor-year which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups and a dominant accident sequence in each accident group was also identified in this study.

This study aims to understand a time trend of the occurrence rate of steam generator (SG) tube leak in the existing sodium-cooled fast reactors (SFRs) based on the observed data. The target on SFRs in the present paper is Phenix in France and BN600 in Russia. From the open literature review, we investigated the number of tube-to-tubeplate weld, the number of tube-to-tube weld, heat transfer area of tube base metal, operating time of SGs, dates when SG tube leak occurred, leaked location, corrective action after tube leak such as replacement of leaked module. Based on these observed data, time to leak is estimated and then time trend of the occurrence rate of SG tube leak for each of the above-mentioned parts was quantitatively analyzed by the hazard plotting method. As a result, the rate of leak at tube-to-tube weld in Phenix shows increase with time due to probable cause of cyclic thermal stress in a short term. As for a long-term trend, the rate of tube leak in both Phenix and BN600 SGs indicated decrease with time probably thanks to improvement in welding and in SG operating condition and to removal of initial failure.

This paper outlines the methodology taken in the preparation and execution of the fire-induced multiple spurious operation (MSO) for the Shimane Unit 2 Nuclear Power Plant, as well as introduce key takeaways. Following the near-catastrophic fire event in the Browns Ferry Nuclear Power Plant in 1975, the US Nuclear Regulatory Commission (NRC) has taken action to require the evaluation of fires that could lead to the simultaneous spurious operation of multiple electrical equipment. In the Fire PRA, the ASME/ANS PRA standard [1] requires the PRA analyst to consider spurious operations that could contribute to an initiating event, affect the functionality of systems credited in the PRA, and/or result in the loss of reactor coolant system integrity. The Shimane Fire PRA is being performed in accordance with the standard, as well as the EPRI/NRC Fire PRA Methodology[2], and thus the applicability of MSO scenarios has been discussed in an expert panel held in June 2019. In this panel, experts on system design, electrical engineering, fire safe shutdown analysis, and PRA, gathered in the office of Hitachi-GE Nuclear Energy in Hitachi-city, Ibaraki, reviewed the compiled list of potential MSO scenarios and discussed their applicability for Shimane Unit 2 over a span of one week. This MSO list consisted of generic potential MSOs listed in NEI 00-01[3], augmented by plant-specific scenarios identified in the pre-panel reviews of piping & instrumentation diagrams and electrical schematic diagrams. The output of this expert panel was a plant-specific list of equipment failures to be added in the FPRA. As a result, the expert panel, over seventy scenarios were judged to be applicable to Shimane Unit 2, as well as additional action items to be considered later. An example of the identified scenarios is the failure of an injection system due to flow diversions caused by fire-induced inadvertent opening of multiple valves. Equipment associated with each applicable MSO scenario were identified and added to the Task 2 equipment list, and their cables were selected as part of the Task 3 cable selection. The event trees/fault trees were also modified to capture the effects of each scenario as part of the Task 5 process.

Sensitivity analysis and uncertainty analysis are used to quantify the influence of each key uncertainty parameter on key output parameters in complex systems, including quantifying the contribution of each input parameter to the uncertainty of the output parameter and the overall uncertainty of the output parameter. Sensitivity analysis and uncertainty include deterministic analysis method and statistical analysis method, among which statistical analysis method is widely used because it has better flexibility in uncertainty analysis. When statistical analysis method is used for sensitivity analysis and uncertainty analysis, it is necessary to generate a sample set by sampling method, and repeatedly call the main calculation program, resulting in high computational cost. Sensitivity analysis and uncertainty analysis based on surrogate models can greatly improve computational efficiency. In this paper, based on the sensitivity analysis of key parameters of effective proliferation factor Keff, a surrogate model based on Kriging interpolation is constructed according to the variation characteristics of the effective proliferation factor under different uncertain engineering parameters such as fuel rod radius and density. The surrogate model based on Kriging interpolation can be used to analyze sensitivity effectively

The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures at ultra-high temperature by using the failure mitigation technology. At the beginning, to identify the accident sequences having the potential to improve resilience, the characteristics of a next-generation loop-type sodium-cooled fast reactor (SFR) in Japan has been investigated by analyzing the event tree of level-1 and level-2 probabilistic risk assessment (PRA). As a result, an event of loss of heat removal systems (LOHRS) is identified. The effectiveness of the measures for improving resilience is evaluated by quantifying the reduction rate of core damage frequency (CDF) before and after the introduction of the measures for improving resilience for all the accident sequences leading to LOHRS. To examine applicability of the developed methodology, a trial evaluation has conducted for a next-generation loop-type SFR in Japan. Through the applicability examined, the method for the effectiveness evaluation was developed successfully. The refinement of the conditional success probability of the measures for improving resilience is the future work.

Various measures have been taken against external events (earthquakes, tsunamis, external fires, tornadoes, etc.), according to new regulations, formulated considering severe accidents caused by the 2011 off the Pacific coast of Tohoku Earthquake. To insure that these measures will function without any problems in the event of an emergency, reliable maintenance management of the facilities themselves, training of personnel for the purpose of confirming the effectiveness of procedures drawn up, as well as confirming that prerequisites for various evaluation methods have not changed are speculated in the Tech. Spec. thus the periodic confirmation of safety and improvement of the skill level of operators is carried out.

In addition, for the purpose of studying effective measures to further improve safety, we have conducted a study of additional measures focusing in a preferential manner on issues that contribute principally to the overall risk, through analyzing results of probabilistic risk assessment including these facilities, measures, etc.

In the recent several years, as the growing demand for energy supply in offshore areas, there is increasing interest in the Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform and would be deployed worldwide. The concept design of OFNP-300/OFNP-100 has been promoted at Massachusetts Institute of Technology, and the Akademik Lomonosov FNPP is ready for operation in Russia. Land-based nuclear power technology has gradually turned to be mature through half a century of development. Nevertheless, it's important to pay attention to the environmental adaptability, the systematic integrity and the equipment reliability in FNPP because of the compact structure, strong maneuverability and lack of support. To reduce the risk in the research and development processing, technology readiness assessment (TRA) is a necessary process to select the best technologies meeting the system requirements by examining maturity of the technology. However, the general method of the TRA only applies to single key technology, it is difficult to assess the integrity readiness and system readiness of nuclear power system in FNPP. In this paper, the method of the TRA based on the integration of Delphi and AHP method be adopted to solve these problems. Using the method developed, the multi-hierarchy key technology system, the assessment model and the assessment process were built. In addition, the low-hierarchy single technology readiness levels, the mid-hierarchy integrity readiness levels and the high - hierarchy system readiness levels of nuclear power system in FNPP were assessed. The result shows that the assessment levels of the single technology readiness of nuclear power system are generally high. Moreover, the assessment levels of the integrity readiness and system readiness of nuclear power system in FNPP are relatively low. As a result, it’s necessary to pay attention to the integrity between the technologies and systems in follow-up research and development process. Furthermore, the method developed are not only be used to assess the TRA of nuclear power system in FNPP, but also can be applied in other complex engineering management field..

After the accident at the Three Mile Island nuclear power plant in the United States, the probabilistic safety assessment (Probabilistic Safety Assessment, referred to as PSA) in the nuclear industry began to boom. Although nuclear power plants have various safety designs, reactor core damage accidents can still occur. Using PSA technology can help nuclear power plants identify weak links and reduce the chance of core damage accidents within a reasonable range. The NRC believes it is necessary to establish a mechanism to monitor the effectiveness of maintenance operations at nuclear power plants to ensure that critical safety systems are able to perform their assigned safety functions. In July 1991, the NRC issued 10 CFR 50.65 Maintenance Rules (MR), titled "Requirements for Monitoring the Effectiveness of Nuclear Power Plant Maintenance." In 1993, the Electric Power Research Institute (EPRI) issued NUMARC 93-01 "Industry Guidelines for Monitoring the Effectiveness of Nuclear Power Plant Maintenance". On July 10, 1996, the maintenance rules came into effect. Effective maintenance closely related to nuclear power plant maintenance and operation and nuclear safety can minimize the number of transient events due to system, structure and component (SSC) failures. To evaluate and/or monitor the effectiveness of maintenance and operations, Maintenance Regulations (MR) use probabilistic safety assessment (PSA) techniques to ensure that each system is performing well. A peer-reviewed PSA model is an appropriate tool for conducting (a)(4) evaluations. In general, risk assessment staff should evaluate the risk of planned maintenance activities (such as preventive maintenance) on the previous day or two and ensure that potential risks are acceptable and controlled. If certain emergencies are likely to change the conditions of a previously (or planned) evaluation performed, PSA staff should re-evaluate the risk due to the change in conditions if it falls within the scope of the MR. Depending on the results of the evaluation, planned maintenance activities may need to be suspended or rescheduled.

With the extensive development of risk-informed application and the gradual implementation of technical policies for configuration risk management, risk monitor has gradually shifted from the original offline operation state to the online operation state, that is, integrated into the production activities of nuclear power plants. The calculation speed of real-time risk is generally required to be within 1-2 minutes; The risk monitor used in domestic and foreign basically meet this requirement. However, in multi-task scenarios, such as a risk assessment of maintenance plan, such speed is still far from meeting the application requirements. In addition, how to update the risk monitor database quickly and efficiently is also a problem faced by the risk monitor. Based on the application scenarios and calculation speed requirements of risk monitor, this paper studies an efficient update method of database, an efficient algorithm of computing minimum cut set(MCS) and supporting multi-task scenarios after risk model reconstruction. Combined the latest computer technology, it improves the calculation speed of risk monitor, meets the speed requirements for efficient calculation of multi-task scenario risk, and provides data algorithm for the next stage of risk monitor development.

An importance for risk management has come to be re-recognized in Japan after the accident of the Fukushima Daiichi nuclear power plants. A Probabilistic Risk Assessment (PRA) fulfills a significantly important role in the framework of the risk management. Technical adequacy of PRA model and high quality of data are required in order to utilize PRA for such riskinformed activities. Therefore, Japanese utilities launched the project for enhancing the quality of PRA models and are upgrading PRA models to comply with international standards.

Chugoku Electric Power Company (Chugoku) is also upgrading the internal events at power PRA model including both the level 1 and level 2 (except the source term analysis) and enhancing PRA quality in order to utilize risk insight for improving nuclear safety. The aim of the PRA upgrade is to reflect international state-of-the-practice approaches and to meet ASME/ANS PRA standard[1] requirements (Capability Category II). Our PRA upgrading process is divided into 3 phases: “Phase I”, “Phase II” and “As-is”. This paper reports the detail of PRA upgrade implemented in Phase I and Phase II such as initiating event refinement, accident sequence re-evaluation, detailed fault tree (FT) analyses, modeling severe accident measures and updating human reliability analysis (HRA).

In order to evaluate nuclear power plant (NPP) safety, the nuclear industry has developed two safety analysis methods, deterministic safety analysis (DSA) and probabilistic risk assessment (PRA). However, carrying out DSA and PRA alone and relatively independently cannot accurately reflect the interactions between failures of equipment and the performance of system, thus safety and safety margin of NPP may be too conservative or under-conservative. In recent years, Riskinformed Safety Margin Characterization (RISMC)" technology was proposed by U.S. RISMC couples DSA and PRA to realistically analyze accident evolution paths, accident consequences, and probabilities under various accident scenarios. This paper aimed to solve the problem of huge calculation amount and computing resources in the RISMC analysis of NPPs. The dynamic event tree (DET) method was used and the concept of adaptive search was introduced, in order to balance the accuracy of limit surface positioning and total calculation amount. A typical accident case study of Small-break LOCA accident was carried out. Its results showed that using Adaptive DET algorithm in RISMC analysis could greatly improve the calculation efficiency of RISMC in accident scenarios and break through the limitations of traditional safety analysis of NPP.

This paper introduces the practical experience of risk-informed technology in China. Taking a pressurized water reactor nuclear power plant in China as an example, aiming at the problems and risks existing in the periodic test one cycle of pilot operated safety valve (SEBIM valve) of Residual Heat Removal (RHR) system of nuclear power plant, and according to the essence of risk-informed technology, a plan for optimizing the periodic test cycle is proposed. This paper mainly discusses the application of risk-informed technology in the periodic test cycle of SEBIM valve in RHR system of nuclear power plant by using the combination of determinism and probability theory. The results show that the extended periodic test cycle of SEBIM valve in RHR system meets the requirements of relevant regulations and guidelines, and the risk introduced by optimization is small and acceptable, which verifies the feasibility of the optimization scheme, and points out the degree of influence of long-term performance monitoring of valves on the optimization scheme, so as to support the comprehensive decision-making of nuclear power plants.

At present, the usual practice of Chinese nuclear power plants is to maintain the Living PSA model and the Risk Monitor model as two independent models. This paper summarizes a set of methods, which can integrate the average model and the instantaneous model into one RISKSPECTRUM file (*.rpp) based on the practical experience of Qinshan NPP (Nuclear Power Plant).

The method can facilitate and simplify the long term maintenance workload of the PSA model. The main points are as the following:

1)The fault tree is used as the input of initiating event of the event tree to process the frequency value corresponding to each Plant Operating State (POS).

2)The method of preprocessing event tree is used to deal with the lumped initiating event, which can well reduce the number of split event trees.

3)Based on the above 1) and 2), two sets of Boundary Condition are used to enable that one RISKSPECTRUM file (*.rpp) can get both RM instantaneous CDF and Living PSA average CDF.

As an important nuclear safety supervision tool, the process of accident sequence precursor (ASP) event evaluation is divided into qualitative screening, quantitative evaluation and statistical analysis. At present, the statistical analysis of ASP events only preliminarily shows the trend of the number or occurrence rate of various types of ASP events over time, and can only rely on the number of events to visually determine major adverse trends, rather than carrying out deterministic analysis of various trends through specific methods. Starting from the ASP events of nuclear power plants, this paper explores and compares different trend analysis methods, focusing on the use of deterministic mathematical methods to carry out the trend analysis of ASP events, so as to identify and determine major adverse trends as soon as possible, and continuously improve the nuclear safety supervision level of power plants with higher quality by combining the quantitative evaluation of ASP events.

With the operating experience and understanding of the nuclear power increasing gradually deeply, the industry believes that it is very necessary to optimize the hydrostatic test cycle of the primary circuit of the nuclear power plant, so as to optimize the hydrostatic test cycle and improve the operational performance factor of the nuclear power plant. This paper proposes a risk assessment method for the change of the hydrostatic test strategy of the primary loop of the nuclear power plant based on the influencing factors of the change of the hydrostatic test strategy，includes implementation process, the technical sufficiency requirements of the PSA model, risk acceptance criteria and other technical elements. Then, taking a domestic power plant as an example, the risk changes of its primary circuit hydraulic test strategy change were analyzed based on the proposed risk analysis method. The research results show that the risk increment because of change of the hydrostatic test strategy of the primary loop of the nuclear power plant still meets the risk acceptance criteria in RG1.174. Thus, the change on the hydrostatic test strategy of the primary loop of the nuclear power plant is acceptable. The risk analysis method proposed in this study can provide a reference for the optimization and change of the hydraulic test strategy of the primary circuit of the same type of nuclear power unit.

It is important to conduct a multi-unit probabilistic risk assessment (MUPRA) to correctly identify the risk of external hazards for a site with multiple units. In Japan, an earthquake is an important external hazard in site risk. In this study, we have developed a method to assess seismic-induced initiating events (SMUIE) frequencies in a multi-unit site.

To develop the method that can utilize insights and results in seismic single-unit probabilistic risk assessment (SUPRA), we have developed a method to estimate SMUIE frequencies by extending an initiating event classification tree method commonly used in Japan.

The proposed method generates SMUIEs by combining seismic initiating events available in seismic SUPRAs. In the SMUIE generation process, the following factors specific to a multi-unit site are considered: (1) failure of shared equipment among units, (2) cascading effects across units, and (3) correlation of response and correlation of capacity of SSCs among units. Then, the SMUIEs are identified by checking validities of the combination considering these factors.

In addition, this paper provides a preliminary evaluation using the proposed method.

To understand the risk induced by tornado missiles, we need to evaluate both the degree of potential damage incurred and the frequency of the occurrence. In the paper, we show two methodologies effective to evaluate the degree of potential damage and the probability of the occurrence of tornado-borne missile strikes on structures, systems, and components in a nuclear power plant. One methodology is based on a tornado wind hazard model TOWLA, and a deterministic tornado-missile analysis code TONBOS, both of which were developed by the CRIEPI researchers including the present authors. The other is based on a probabilistic tornado-missile analysis code TONBOS-pro, and an annual missile-strike probability evaluation code TOMAXI-pro, both of which were developed by the first and third authors. To concretely show the effectiveness of the proposed methods, the annual probability that trucks on a straight road collide with a structure was evaluated by the two methods. The numerical results were compared with each other to indirectly confirm the validity of both evaluation methods. As a result, it turned out that the annual strike frequency computed by the deterministic evaluation method was much higher than that of the probabilistic evaluation method. Since the former method includes conservative assumptions, the results are consistent with the evaluation presumptions, reasonably indicating the validity of both methods.

The Distributed Control System (DCS) of the nuclear power plant plays an important role as the operation center that realizes the process control as well as the centralized monitoring and management under all different operating conditions for the entire nuclear power plant. Safe, reliable and effective operation of DCS is considered to be a pivotal guarantee for the design concepts of nuclear safety, e.g., defense in depth of the nuclear power plant, which is very critical to the safety and economy of nuclear power plants. However, the centralized management of DCS leads to a great increase in the possibility that the multilayer preset barriers are destroyed by one local fire in the nuclear power plant, eventually threatening the safety of the power plant. Different from the other systems in the nuclear power plant, the DCS is mainly composed of different components like computers and servers with sophisticated structure, which involves transmission of complex logic signals. These make the fire risk assessment of DCS quite challenging. In this paper, by considering the DCS in the Chinese 2nd-generation improved unit of the pressurized water reactor (PWR), we propose a method of fire risk analysis for DCS, based on probability theory and system reliability evaluation technology. By establishing the logical relationship between the spatial location of fire in the plants and DCS racks, cabinets, optical cables, signals and related equipment, an analysis method is proposed for the possible initiating events after a fire occurs in the DCS-related cabinets. Then a modeling strategy is developed for the impacts of the DCS fire failure on the accident mitigation function. The proposed method of fire risk analysis for DCS is capable of achieving quantitative assessment under all different operating conditions of the entire plant, which, thus, can support comprehensive PSA analysis of the internal fire.

The probabilistic safety analysis of nuclear power plant shows that the emergency diesel generator has a significant contribution to the safety of the plant, especially the common cause failure. In order to reflect the design characteristics and operation experiences of domestic nuclear power units, the paper takes the emergency diesel generator of domestic VVER nuclear power units as a research object, uses the event impact vector analysis method to develop the specific data analysis of common cause failure. The paper provides a method to estimate common cause failure model parameters based on the generic data, design characteristics and operation experiences of target plant. On this basis, the sensitivity analysis of the applicability factors involved in the method is carried out.

Risk informed in-service inspection (RI-ISI) is one of the most successful applications of probabilistic safety analysis techniques, RI-ISI has been implemented in a large number of operating power plants at home and abroad, Under the premise of ensuring safety, the dose of radiation for staff is reduced for nuclear power plants, reducing the economic burden. However, RI-ISI is a gap in the design phase of nuclear power plants. In this paper, the method of RI-ISI optimization of nuclear power plants at the design stage is studied, and the pipeline in-service inspection scheme of HPR1000 TFA/TFM/PRS/PCS system is optimized by using the RI-ISI method. Through qualitative analysis and quantitative calculation of RI-ISI risk consequences, the in-service inspection optimization scheme of the studied system is given, which effectively ensures the safety of nuclear power plants while reducing the task volume and time cost of in-service inspection work, effectively reducing the irradiation dose of in-service inspection staff, and bringing greater economic benefits to nuclear power plants.

Level 2 PSA could give us the quantitative results of large release frequency (LRF), containment response and source terms. Moreover, it is essential for the demonstration on safety goals and Emergency Plan Zone (EPZ). ACP100 is an integrated SMR design developed by China, which adapts the advanced passive safety system and innovative design philosophy. In this study, it is focused on the study of ACP100 full-scope Level 2 PSA, which consists of internal and external events, as well as full power, low power and shutdown conditions. In this study, firstly, it will introduce the related requirements and ACP100 design features related to severe accident phenomenon and management. Secondly, it will summarize some key techniques of ACP100 full-scope Level 2 PSA study process. Finally, it will analyze the internal flood Level 2 PSA study case as an example with conservative assumptions, discuss the conclusion and outlook.