This paper aims at verifying the current Japanese Emergency Response Guideline, especially the notification scheme of Emergency Action Level (EAL), through the analysis of the progress of Fukushima nuclear accident. We compared timing and emergency classification between two data sets of the plant statuses which expressed by the old prediction-based notification and the latest EAL-based notification, in order to assess the current EAL scheme along the effectiveness of protective action for the local residents. We observed that the plant statuses expressed by the current EAL-based notification gave more engineering insights in the earliest accident phase. We also identified a potential improvement area of the guideline in the following severe accident management (SAM) phase after the trigger of first precautionary action, where we are required to reduce uncertainties in both processes of the operator’s notification and the government’s decision-making, in order to compensate for the abstention of utilizing predictive information.

Nuclear accident and nuclear terrorism are of great threats to international security. The countermeasures against nuclear emergency situation should be deliberately considered before it happened. In accordance with experience from Fukushima Daiichi Accident, various kinds of complex situations must be considered at the phase of emergency preparedness. And all related schedules, including national laws and regulations, allocation of responsibility and authority, emergency plans and procedures, facilities, instruments, trainings and exercises must be examined. The appropriateness of the emergency preparedness and the emergency response capability need to be verified through emergency exercises. However, exercises are usually costly and labor-intensive. So, it is necessary to design a software system to help conduct emergency exercises simulation. A Nuclear Emergency Decision Deduction and Training Platform (NEDDTP) is designed in China Nuclear Security Center of Excellence, which aims at performing the simulation of nuclear emergency process and providing a training platform for emergency workers. 3D models, the environment of nuclear power plant (NPP), some physical numerical models for simulating and several typical nuclear emergency scenarios for training are included in the platform. NEDDTP realizes the function of digitization of scenario, emergency simulation, resource management, command and dispatch, process record and evaluation in exercise, and it also provide VR display for emergency workers.

Determination of the SMR emergency planning zone (EPZ) is one of the important external constraint factor of its marketing and application, which means that it is very important to formulate appropriate classification criteria and establish proper size range. In China, due to the requirement of “Criteria for emergency planning and preparedness for nuclear power plants: Part 1, The dividing of emergency planning zone.” (GB/T 17680.1-2008), for PWR nuclear power plant, its external plume EPZ should be within 7km–10km, and its internal plume EPZ should be within 3km∼5km. However, the scope of the standard for the emergency planning area is currently limited to conventional nuclear power plants, and for the current SMR, its emergency planning size is not included. In this paper, we will analyze the classification method of SMR EPZ based on the traditional Nuclear Power Plants feedback experience, including selection of source term, accident cutoff probability, determination method of the plume EPZ and the ingestion EPZ. Three typical nuclear power plant sites in China are chosen as CAP200 case study sites, including two inland nuclear power plant sites and one coastal site. The three sites can represent most of the meteorological and terrain characters of China nuclear power plants. According to the CAP200 source term and meteorological data of the sites, MACCS2 computer program is used to calculate the severe accidents consequence. Conclusions show that for the CAP200 SMR, the accident cutoff probability can be 1.0E−08 to 1.0E−07 per reactor per year, and its project dose exceeding probability in the three sites boundary is far below 30%, which directs that for CAP200 SMR, its plume and ingestion emergence planning zone is limited to the on-site area, and its off-site emergency response can be simplified.

In a nuclear accident, radioactive release source term is the critical factor of nuclear emergency response and accident assessment. The modelling of source inversion based on variational data assimilation (VAR) is capable of balancing the environmental radioactive monitoring data to obtain the global optimal source term. But it could be influenced by the discrepancy between predictions of the atmospheric dispersion model and observations, which is defined as the dispersion model error in this study. In order to reduce this influence, the VAR with the dispersion model error (DME-VAR) is proposed. In the DME-VAR, the dispersion model error is quantified by the error coefficients at every monitoring station. These error coefficients and the release source term are estimated at the same time. For limiting the runtime, the DME-VAR program supports parallel processing. Two sets of wind tunnel experiment data for a typical Chinese nuclear power plant site are used to validate and evaluated the performance of the DME-VAR. The results demonstrate that the DME-VAR effectively estimates the error coefficients, and outperforms the VAR in both release rate estimation and radioactive contamination predicting. Moreover, the runtimes of these verification experiments are all reasonable, even for the application in the nuclear emergency response.

An inverse source estimation method is proposed to reconstruct emission rates of multi-radionuclides using local gamma dose rate measurements under the data assimilation framework. It involves the Proper Orthogonal Decomposition (POD)-based ensemble four-dimensional variational data assimilation (PODEn4DVar) algorithm and a transfer coefficient matrix (TCM) created using FLEXPART, a Lagrangian atmospheric dispersion model. PODEn4DVar is a hybrid data assimilation method that exploits the strengths of both the ensemble Kalman filter (EnKF) and the 4DVar assimilation method. With an explicit expression of control (state) variables in the cost functional, the data assimilation process is substantially simplified than traditional 4D variational method. By setting a unit emission rate and running the ATDM model (FLEXPART in this article) driven by meteorological fields forecasted with WRF, we get the transfer coefficient matrix with the progression of nuclear accident. TCM not only acts as observation operator in PODEn4DVar, but also eliminates the control run in traditional data assimilation framework. The method is tested by twin experiments with ratios of nuclides assumed to be known. With pseudo observations based on Fukushima Daiichi nuclear power plant (FDNPP) accident, most of the emission rates were estimated accurately, except under conditions when wind blew off land toward the sea and at extremely slow wind speeds near the FDNPP. Because of the long duration of accident and variability of meteorological fields, measurements from land only in local area is unable to offer enough information to support emergency response. With abundant measurements of gamma dose rate, emission rates can be reconstructed sequentially with the progression of nuclear accident. Therefore, the proposed method has the potential to be applied to nuclear emergency response after improvement.

Radiation monitoring instruments (KRT) is important to decide emergency response level (EAL) in accident situation. Emergency response drills is more and more significant after Fukushima Daiichi severe accident. This article develops a code to simulate radiation monitoring instruments data for emergency response exercises. A part of calculated input of the code comes from MAAP calculated results included source terms and thermo hydraulic data. KRT simulation code runs with MAAP calculation results and output KRT simulated value at the same time. This article gives a way to prove the result of KRT simulation code analyzed is correct and matched through simulating an emergency response exercise scenario.

Fukushima Daiichi accident shows that extreme external events may lead to extensive damage in nuclear power plants (NPPs), which may not only affects equipment, but also damage the control and monitor ability of control rooms (including human factor, habitability, control panel and so on). Accidents-related procedures of NPPs in China are Emergency Operation Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs), both of them rely on control and monitor ability of control rooms. Therefore, current structure of accident management procedures can’t work effectively in the scene of extensive damage. The development of Extensive Damage Mitigation Guidance (EDMGs) may strengthen the defense in depth. In relation to this, an Integrated EDMG has been developed to deal with extensive damage resulted from fire (or explosion) or extreme hazards, while the Nuclear Energy Institute (NEI) EDMG only considers fire and explosion. Besides the extended scope, Integrated EDMG develops a comprehensive and detailed mitigation method which is applicable for extensive damage. In this paper, the characteristics of extensive damage will be analyzed first, and then the structure of Integrated EDMG is developed based on the characteristic analysis. According to the emergency response analysis and accident management research, detailed introduction of the Integrated EDMG and its sub-guidances are discussed.

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3 rd generation nuclear power plant, which applies the advanced European nuclear power technology. This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions. Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

With the introduction of third generation of nuclear power AP1000, Westinghouse uses the mobile device (a mobile wastewater treatment device 6 units shared) radioactive waste system design concepts. This design not only simplifies the process of nuclear island waste system; saves equipment layout space; improves equipment utilization; while increases the use of new technologies lifetime of the plant and the possibility of flexibility. This paper introduces the first AP1000 unit (Sanmen, Zhejiang Province) by using the advanced mobile device technology and application of wastewater treatment under the condition of the primary coolant source level. At the same time, the paper also discusses the periodic system inspection and the strategy of maintenance. In addition, the paper further expands the application direction of the mobile waste processing aspects, such as: decommissioning of nuclear facilities; enhancing the facility decommissioning radioactive liquid waste purification capability. Another example: After the Fukushima accident, people pay more attention to accident-mitigation-design and hope to accelerate the development of emergency radioactive liquid waste processing devices. Thus, in addition to strengthening the nuclear power plant inherent defense in depth and resistance emergency capability, mobile waste treatment device or combination device special regional settings can be made to improve and enhance the ability to get more diversified emergency response.

Nuclear emergency is the important component part of the defense in depth philosophy of nuclear safety, which is the last barrier to protect public in a nuclear accident. According to Chinese codes and standards which are Regulations on the Nuclear Accident in Nuclear Power Plant and Emergency Preparedness and Emergency Response of the Research Reactor (HAD002/06), the research reactor operation units should carry out a comprehensive emergency drill at least once every two years, in order to check the effect of the emergency training about the on-site emergency organization and verify the usefulness and feasibility of the emergency plan. Therefore, the Institute of Nuclear and New Energy Technology (INET), Tsinghua University, had drew up the comprehensive emergency plan, conducted the emergency practice in December 2015, and tested the emergency preparedness of multiple research nuclear installations including the 10MW high temperature gas-cooled reactor (HTR-10), 5MW test heating reactor, shielding-experimental reactor, etc. The purpose, accident scenarios, process, and key content of the emergency drill were descripted in details, the design of the guiding worksheets which were adopted in the emergency drill of INET for the first time were distinctly introduced, and the effect of these guiding worksheets was explicitly discussed. The whole process of the drill in INET in 2015 was reviewed and summarized, the experience feedback of which can supply useful reference information and recommendations for the development of the emergency drill of research reactors.

In case of a severe accident, all the staff members of a nuclear power plant (NPP), members of the Emergency Response Organization (ERO) on site as well as operators in the main control room (MCR) are required to take necessary actions to mitigate the consequential effects of the accident. Therefore, Nuclear Engineering Ltd (NEL) has been implementing education and exercise for severe accident (SA) management both at nuclear power plants and Nuclear Power Division of Kansai Electric Power Company (KANSAI) since FY 2014. For the education of commanders who take a lead at the ERO in case of an accident, table top exercise is provided by using simulators developed by NEL, including functions to respond to a SA involving core melt. Continuous implementation of this education and exercise program is expected to enhance KANSAI’s severe accident management ability and their voluntary safety improvement activities in the future.

To ensure that adequate protective actions are in place for the public, a salient lesson learned from Fukushima is the necessity to improve the effectiveness of the off-site response, namely the effective implementation of protective actions in a nuclear emergency. Among recent research on nuclear emergency, little attention has been paid to public participation, where the disconnect between the public and authorities, and its negative effect on emergency response exist. This study conducted an analytic discussion on the effectiveness of off-site nuclear emergency, from the standpoint of public participation. The two key factors contributing to effective emergency responses in a nuclear emergency were identified to be the feasibility of emergency plans and the adequacy of emergency preparedness (EP), to which the passive role the public has been playing does no good. First, nuclear emergency plans are developed unilaterally by emergency managers and authorities, without the public involved. This government-centric planning process usually fails to meet the actual needs of the residents should a nuclear accident occur, consequently impairing the feasibility of emergency plans. As regards EP, emergency management’s efforts have long been dedicated to maintain the response capabilities of emergency response personnel, while overlooking the EP of the public. In this case, the public will not be well-prepared for an emergency. Corresponding to the deficiencies stated above, possible solutions to improve the overall effectiveness of off-site emergency response were proposed, from the perspective of increasing public participation. First, to make emergency plans feasible and comprehensive, 1) the public can be incorporated in planning process to consider their needs in emergency plans, 2) emergency plans should be periodically assessed and updated accordingly, based on the up-to-date socio-demographic information. Second, to ensure the effective implementation of EP, 1) the public should be educated more on the knowledge of radiation protection and emergency response, in a participatory rather than informational way, 2) More-realistic nuclear exercises, such as evacuation drills of the population-at-risk, could be cautiously carried out, to test whether the public are well-prepared under emergency conditions. Finally, a precondition of broad public participation is that the public have interest in nuclear emergency. To this end, information communication technologies, should be widely utilized in nuclear emergency to generate public interest, by facilitating two-way communication and displaying the emergency-related information in an easy-to-read way. This study indicates that nuclear emergency should not be a process dominated by emergency managers alone, since the public are not only the protected but also the true first responders in nuclear accidents. Wider public participation should be incorporated into the whole process of emergency management, from planning to preparedness, to maximize the effectiveness of the off-site response to a nuclear emergency.

Core damage assessment is of great importance to the emergency response of nuclear power plants. In this paper, the neural network method is introduced into the core damage assessment process. The hydrogen concentration, together with the temperature and pressure in the containment, are taken as the input parameters of the model. With the simulated result of MAAP codes as the sample data, a neural network model is developed to reconstruct the core overheating damage fraction. According to the calculation of the neural network model, the deviations of the reconstructed results are quite small compared with the simulation results, and one of the typical errors is 1.76%. It can be concluded that the model based on neural network method satisfies the analysis accuracy requirements and can be used as a diverse analytical tool in the core damage assessment of nuclear power plant.

The source term information of radioactive release in a nuclear accident is important for nuclear accident classification, radiological consequences evaluation and emergency response. Two major categories of source term estimation techniques are forward method based on the status data of the nuclear reactor and backward method based on environmental monitoring data. Although the forward method is more widely used, it may introduce large uncertainties into the source term estimate due to its subjective parameters and low quality of reactor status data in a severe accident. To reduce these uncertainties, a coupled source term estimation method that combines both forward and backward models, is proposed in this study. The forward part provides a source term forecast model based on Response Technical Manual RTM-96. Under the framework of sequential data assimilation, the backward part iteratively reduces the uncertainties in the source term estimate using environmental monitoring data. Numerical experiments with different uncertainties are performed using the reactor and monitoring point information of real Chinese nuclear power plant and its atmospheric tracer experiments.

The accident at the three reactor units at Fukushima Daiichi showed weaknesses in the plant coping capability for beyond design basis accidents caused by extreme external events. The weaknesses included plant design features, accident management procedures and guidance, and offsite emergency response. As a result, significant changes to plant coping capability have been made to light water reactors worldwide to enhance the coping capabilities for beyond design basis accidents. However, the response in the United States has been significantly different from that in Europe in a number of ways. In the United States, the regulator and the industry convened separate expert panels to review the Fukushima accident and make recommendations for enhancements. On the regulatory side, a series of three Orders were issued and that required the implementation of certain enhancements (Mitigation strategies, hardened vents for certain BWRs, spent fuel pool level indication) to ensure adequate protection for the health and safety of the public. Other enhancements were subject to the “Backfit Rule” which requires that changes to regulatory requirements be shown to be cost beneficial using accepted methodologies. Simultaneously, the industry took independent steps to develop a diverse and flexible coping strategies (known as FLEX) and other enhancements. The focus in the United States was clearly on enhancements to guarantee continued core, containment and spent fuel pool cooling in the event of beyond design basis accidents, particularly those resulting from extreme external events. In Europe, the regulatory agencies ordered the development and completion of “Stress Tests” for each reactor site. These Stress Tests were focused on identifying the capability of the plant and its staff to respond to increasingly severe external events. The Stress Tests not only examined the ability to maintain core, containment and spent fuel pool cooling but also the ability to mitigate the consequences of accidents that progress to core damage (i.e., a severe accident). Regulatory requirements were then issued by the national regulators that addressed the weaknesses identified from the Stress Tests. While many of the enhancements to the plant coping capability were similar to those in the United States, significant hardware enhancements were also required to reduce the consequences of core damage accidents including hydrogen control and containment filtered venting. Finally, most European regulators also include severe accident management guidance (SAMG) as a regulatory requirement. In the United States, SAMG will be maintained as a voluntary industry commitment that is subject to regulatory oversight review.

Filtered containment venting system (FCVS) is an emergency response system for severe accident. FCVS reduces the pressure in reactor vessel and removes radioactive iodine and cesium. Recently, silver doped zeolite (AgX) has been attracting attentions since it is believed to remove over 99.99% of CH 3 I in the ventilation gas. However, the reliable data of AgX for CH 3 I sorption under expected severe accident conditions are still limited. Therefore, the sorption behavior of CH 3 I on AgX was investigated under several expected conditions. In this study, the CH 3 I labeled by 125 I tracer was synthesized by isotope exchange reaction, and it was devoted for sorption experiment on AgX. Sorption experiment using stable CH 3 I was also conducted by the same procedure. From results of the sorption experiment, over 99.9% of 2.28g of CH 3 I were sorbed by 10g of AgX. After the sorption experiment, AgX samples were analyzed by powder X-ray diffraction (XRD) and SEM–EDX. These results indicated the distribution of Ag agreed with that of I in micro area of AgX. For the evaluation of apparent sorption capacity of AgX, breaking through experiment was carried out. The apparent sorption capacity of AgX for CH 3 I was determined as 0.25g / g (AgX) at 24°C. The breaking through experiment was also carried out for the evaluation of CH 3 I flux and humidity effects in the vent gas. The effect of temperature on the sorption capacity was also studied. When the flux of CH 3 I supply was increased from 1.09 × 10 −4 to 2.33 × 10 −4 g/cm 2 s, apparent sorption capacity was unchanged (= 0.21g / g (AgX)). Besides, the apparent sorption capacity was unchanged in the humidity range of ∼0 % H 2 O to 0.01 % H 2 O, while it decreased to 0.19 g / g (AgX) at 3% H 2 O. When sorption temperature increased to 170 °C, the apparent sorption capacity also increased to 0.69 g / g (AgX).

The Risø Mesoscale PUFF model (RIMPUFF) predicts dispersion of released hazardous material and plays an important role in the emergency response system of many Chinese nuclear power plants (NPPs). The wind field over the calculation domain is a critical input of RIMPUFF, which dominates the performance of RIMPUFF. Due to the complex topography of most Chinese NPP sites, it remains challenging to provide a refined wind field for RIMPUFF prediction. To solve the problem, California Meteorological Model (CALMET) is coupled with RIMPUFF for wind field calculation in this study. Moreover, the computing scope and capability of RIMPUFF are enhanced in order to obtain more accurate prediction for the emergency response and source term inversion in nuclear accidents. Except for that the unlimited amount of grid with higher resolution is supported, a new sampling module is added to RIMPUFF for predicting the concentration of radioactive materials and dose at number-unbound arbitrary location. To verify the CALMET-RIMPUFF method, a wind tunnel experiment that replicates the topography of one Chinese NPP site within 10 km-range, is conducted. In the experiment scenario, the speed and vertical profile of the incoming flow is carefully set according to the annual mean wind speed and wind profile data measured in recent years on the meteorological tower of this NPP. The results demonstrate that the wind field calculated by CALMET is consistent with the topography. With this wind field, the RIMPUFF-predicted concentration distribution matches the measurements well both qualitatively and quantitatively. Moreover, the calculation of U.S. Environmental Protection Agency (EPA) statistical evaluation metrics indicate that the random scatter is within a factor of 1.8 and the FAC2 is nearly 80%. It proves the acceptability of CALMET-RIMPUFF over the complex topography of Chinese NPP sites.

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

The source information of the radionuclide release in nuclear accidents is a key issue of the nuclear emergency response. One way to estimate the source information is by inversing the radionuclide transportation process based on environment radiation monitoring data. The advantage of this method is that the required data are easy to obtain in accident. But it is vulnerable to large uncertainties in both data and transport model. To solve the problem, a source term estimation method based on four-dimensional variational (4DVAR) data assimilation technique was proposed for source term estimation in this study. The proposed method couples 4DVAR with the RIMPUFF air dispersion model. It formulates the inverse modelling of source term estimation as an optimization problem that is trying to find an optimal balance between real observation data and the background field. The advantage of this method is that the radionuclide transport in every time step is included in data assimilation and the result is global optimum in the whole assimilation period. The gradient for cost function is calculated by the backward integration of the adjoint model. Practical imperfectness of measurement were considered and integrated into the cost function. The proposed method was verified using numerical simulation for both homogeneous and heterologous atmospheric condition. The performance of the source term estimation method was also investigated with respect to different release profile, wind speed and atmospheric stability class. The simulation results demonstrate that the estimate matches the true release status well for both homogeneous and heterologous wind field. Also, the experimental results show that the proposed method has strong robustness to wind speed and atmospheric stability.

In nuclear accidents, source estimation methods can be used to assess the amount and status of radioactive materials emission, which is important for nuclear emergency response. Current studies primarily focus on source estimation of a constant release mode, which may not match real nuclear accident scenario. To solve this problem, a fast on-line source term estimation method for non-constant release mode was proposed in this study. Based on the well-established RIMPUFF dispersion model, the extended Kalman filter algorithm was utilized to perform an online estimation of the total amount and height of the release source simultaneously. To simulate the accidental radioactive release mode, a non-constant function was used as the state description of the release source. Both air dispersion model error and measurement error were considered. To demonstrate the feasibility of the proposed method, twin experiments were performed with release profiles. A sensitivity study was also performed to assess the robustness of the algorithm. The experiment results demonstrate that the proposed method can provide robust source term estimation for both constant and non-constant release. The estimation accuracy depends on the steepness of the curve in the growth of the release mode and the distribution of simulated monitors.