Estimation by analysis


  1. To analyze and estimate the location, amount, properties of fuel debris and the FP distribution as they serve as basic information for studying the fuel debris retrieval methods and development of equipment and facilities.

Current status

The amount, locations and composition of fuel debris and FP distribution are to be estimated using severe accident analysis codes. Developing a model specific to each analysis code is to be based on the data obtained from the zircaloy oxidation tests and melting tests of uranium oxide in the past. The result of analysis is to be obtained according to the progress of the accident and scenario, such as the amount of injected water and opening and closing of SR valve.

The severe accident progression analysis is highly depending on the computational model employed and estimation scenario and calculated results contain the uncertainties. However, quantitative information, such as the amount and composition of fuel debris and FP distribution at several locations inside the reactors will be able to be obtained and it will be effective method to understand a whole situation of the severe accident. Also, estimating the temperature history inside the reactors during the progression of the severe accident through the severe accident progression analysis, the states of the major internal structures and equipment is to be presumed by the calculation results.

Furthermore, OECD/NEA BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) project has been carried out as an international joint research. In this project, the estimation of internal PCV conditions is being performed through the severe accident progression analysis by 13 institutions from Japan and abroad.

The estimation by analysis situation performed to date is described below.

(1) Analysis results using severe accident analysis codes

Using the MAAP and SAMPSON codes,which are the severe accident analysis codes, the amount and locations of the fuel debris and FP distribution are estimated. Addition and improvement of physical phenomena model (NDF Technical Strategic Plan 2016,page A-21)required to evaluate the fuel debris and FPs at the Fukushima Daiichi has been performed for both codes. These improvements were completed in FY2015. The amount and locations of fuel debris and distributions of FP were analyzed using the improved version of the MAAP and SAMPSON codes.

Also, in FY2015, the analyses focusing on the event specific to each Unit were conducted. The analysis addresses the clarification of the mechanism of the plant behavior using sensitivity analysis, and reduction of the uncertainties in the analysis. For example, since pressure increase behavior (pressure spikes) was observed three times in Unit 2 after the depressurization of RPV, this pressure behavior was reproduced by the steam and hydrogen and water vapor generated by the reaction of the fuel debris and injected water using the MAAP code(NDF Technical Strategic Plan 2016,page A-22).

The MCCI (Molten Core Concrete Interaction) evaluation implies that most of the fuel debris was highly likely to have fallen to the pedestal in Unit 1 due to the damage to the RPV. It is important to evaluate the erosion of the concrete and amount of MCCI. Since the shape of the pedestal of the Fukushima Daiichi NPS is complicated including sump pits, relocation and diffusion model for the concrete are added to the MCCI evaluation module of the SAMPSON code and the extent of the scattering and erosion behavior of the fuel debris in Unit 1 were evaluated. The evaluation results (NDF Technical Strategic Plan 2016,page A-23)suggest that the fuel debris are scattered toa fairly wide range of the D/W floor of Unit 1.

The results of the analysis of the amount and locations of fuel debris and FP distribution using the MAAP and SAMPSON codes are described below. Also, the estimation results of the conditions of the reactor internals and equipment based on the temperature estimated by the severe accident progression analysis are as follows:

  1. Amount and locations of the fuel debris
  2. The results of analysis of the amount and locations of the fuel debris are shown in Table 1. Since severe accident analysis codes have characteristics of the models used for each code and uncertainties in the input scenario, it is required to take into account the uncertainties contained for use the results. Issues to be noted in the comparison among the result of severe accident analysis codes and Units are show in Table 1.

    Table 1縲Analysis results using the severe accident analysis code (Unit: ton)
    Locations Component Unit 1 Unit 2 Unit 3
    Core region 0 0 0 13 0 29
    Bottom of the RPV 15 10 25 58 25 79
    Inside the pedestal Fuel and structural materials 109 79 92 76 103 53
    Concrete 78 130 37 14 51 20
    Outside the pedestal Fuel and structural materials 33 52 102 5 96 0
    Concrete 52 0 4 0 6 0
    Total amount(concrete included) 287 271 260 166 281 181

    Note: The weight inside and outside the pedestal is the weight of fuels/structural materials (excluding the weight of concrete). The weight of concrete is indicated in ( ).

    シサReferenceシ IRID Completion report for "Improvement of recognition regarding the internal PCV condition using severe accident progression analysis and actual plant data"シス

  3. FP distribution
  4. The FP distribution inside the RPV, PCV and R/B were analyzed by the MAAP and SAMPSON codes.The results indicated a large difference which was depended on properties of FP nuclides using both codes. The results of analysis of the distributions of Cs and Sr, which are the representative FP nuclides, are described in this page . A large difference (uncertainties) between those codes is caused by the differences in FP evaluation models and chemical form of the FP nuclides which were considered in the evaluation model.

  5. Estimation for the status of major structures and equipment
  6. Although it is necessary to know the current conditions of equipment inside the reactor for the fuel debris retrieval, there is no measured value of the environment (temperature) that the equipment have experienced during the severe accident. Therefore, the conditions of the equipment inside the reactor were estimated based on the temperature evaluation results from the severe accident progression analysis (MAAP and SAMPSON codes). This estimation was performed by referencing not only the results of analysis but also the on-site situation. The high-temperature deformation, creep rupture and corrosion degradation were considered as degradation events induced in the structures and equipment in the scope of the evaluation. The Evaluation criteria for each degradation event are shown in this page.(NDF Technical Strategic Plan 2016,page A-25)

    The evaluation results indicated that the creep deformation may be induced for the steam dryer, steam separator assemblies, upper grid plates and core support plates in all Units. For use the evaluation results, the uncertainties are required to be considered in the severe accident progression analysis (MAAP and SAMPSON codes) performed for this time. All estimations results including those of other reactor internals are shown in this page. (NDF Technical Strategic Plan 2016,page A-26)

    Since the evaluation results of the temperature refers to those of FY2014, it is necessary to confirm the effects from the latest analysis results in the future.

    Also,considering the thermal data obtained from the analysis after the severe accident base on the corrosion rate for 40 years after the severe accident, the seismic stress evaluation was performed in the "Development of technology for RPV/PCV integrity evaluation.窶 The results of the analysis indicated that the induced stress at the RPV, PCV and pedestal of Unit 2 fell below the evaluation criteria.

(2) Results of analysis by OECD/NEA BSAF

In the Phase-1 of OECD/NEA BSAF project, the severe accident progression analysis of Units 1-3 for six days after the earthquake was performed by 13 institutions in Japan and abroad. The result of analysis is shown in Table 2.

The result of analysis of Units 2 and 3 was categorized into two cases, which are that the fuel debris remains in the RPV, and falls down to the PCV. The result of Unit 2 can be considered depending on the modeling of the fuel debris relocation from the reactor core to the lower plenum and the assumption amount of the water injection by the fire engine which has large uncertainties. With regard to Unit 3, the results were affected by the difference in the assumption of HPCI water injection behavior (amount of injected water in reducing the RPV pressure), that is, the maximum quantity and cycle of the steam flow that drives HPCI were different significantly among the institutions that performed the analysis.

The Phase-2 of the OECD/NEA BSAF project is being carried out following the Phase-1. In the Phase-2,setting the implementation period at three years, which is from April 2015 to March 2018, the severe accident progression analysis of three weeks after the earthquake is being performed by 22 institutions from 11 countries to advance the improvement of the analyses. Also, the findings of the severe accident research analysis are shared through the workshop regarding the state of the FP adhered to the reactor internals and features of MCCI. The progress (PRG, Program Review Group) meeting and workshop are being held about twice a year. Interim reports and final reports are planned to be established at the end of FY2016 and in May 2018 respectively.

Table 2縲Evaluation results of fuel debris distribution in BSAF Phase-1 (Unit: ton)

Integrating the knowledge around the world, OECD/NEA SAREF (Safety Research Opportunities Post-Fukushima) also studies the project regarding the decommissioning and safety assessment.