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JAEA Reports

Development of burnup/depletion calculation code based on ORIGEN2 cross-section libraries and Chebyshev rational approximation method, CRAMO

Yokoyama, Kenji; Jin, Tomoyuki*

JAEA-Data/Code 2021-001, 47 Pages, 2021/03

JAEA-Data-Code-2021-001.pdf:1.85MB

A new burnup/depletion calculation code, CRAMO, was developed by combining an ORIGEN2 cross-section library set, ORLIB, based on Japanese evaluated nuclear data library, JENDL, and a burnup/depletion solver based on Chebyshev rational approximation method. CRAMO uses the ORIGEN2 cross-section library set ORLIBJ40 based on JENDL-4.0, and the burnup/depletion solver implemented in the versatile reactor analysis code system, MARBLE. It was confirmed that results of CRAMO agreed well with those of ORIGEN2 for burnup/depletion and radioactivity calculation cases. The development of CRAMO made it possible to use ORLIB without using ORIGEN2. It will be possible to provide an easy-to-use processed JENDL data set for burnup/depletion and radioactivity calculations in combination with a burnup/depletion based on Chebyshev rational approximation method. The present version of CRAMO is a subset of ORIGEN2 and can compute only compositions and radioactivities after irradiation. However, since various kinds of outputs of ORIGEN2 can be evaluated by using the composition, it is possible to reproduce many functions of ORIGEN2 by adding post-processing modules.

JAEA Reports

A Numerical simulation study of the desaturation and oxygen infusion into the sedimentary rock around the tunnel in the Horonobe Underground Research Laboratory

Miyakawa, Kazuya; Aoyagi, Kazuhei; Akaki, Toshifumi*; Yamamoto, Hajime*

JAEA-Data/Code 2021-002, 26 Pages, 2021/05

JAEA-Data-Code-2021-002.pdf:2.14MB
JAEA-Data-Code-2021-002-appendix(CD-ROM).zip:40.99MB

Investigations employing numerical simulation have been conducted to study the mechanisms of desaturation and oxygen infusion into sedimentary formations. By mimicking the conditions of the Horonobe underground research laboratory, numerical simulations aided geoscientific investigation of the effects of dissolved gas content and rock permeability on the desaturation (Miyakawa et al., 2019) and mechanisms of oxygen intrusion into the host rock (Miyakawa et al., 2021). These simulations calculated multi-phase flow, including flows of groundwater and exsolved gas, and conducted sensitivity analysis changing the dissolved gas content, rock permeability, and humidity at the gallery wall. Only the most important results from these simulations have been reported previously, because of publishers' space limitations. Hence, in order to provide basic data for understanding the mechanisms of desaturation and oxygen infusion into rock, all data for 27 output parameters (e.g., advective fluxes of heat, gas, and water, diffusive fluxes of water, CH$$_{4}$$, CO$$_{2}$$, O$$_{2}$$, and N$$_{2}$$, saturation degree, water pressure, and mass fraction of each component) over a modeling period of 100 years are presented here.

JAEA Reports

Data of groundwater chemistry obtained in the Horonobe Underground Research Laboratory Project (FY2020)

Miyakawa, Kazuya

JAEA-Data/Code 2021-003, 25 Pages, 2021/05

JAEA-Data-Code-2021-003.pdf:1.91MB

Development of technologies to investigate properties of deep geological environment and model development of geological environment have been pursued in "Geoscientific Research" in the Horonobe Underground Research Laboratory (HURL) project. In the fiscal year 2020, to proceed remaining important issues which were deduced from the conclusion of the investigations during the fiscal year 2015-2019, basic data such as groundwater chemistry need to be successively acquired. In the fiscal year 2020, groundwater was sampled from boreholes drilled in the 140 m, 250 m, 350 m gallery in the HURL, and water rings settled in three each vertical shaft, and groundwater chemistries of 41 samples were analyzed. Here, analytical results of groundwater chemistry such as physicochemical parameters, dissolved ions, oxygen and hydrogen isotope ratios, and tritium content, which were obtained in the fiscal year 2020, were reported along with a detailed description of analytical methods.

JAEA Reports

Ocean current data obtained by Acoustic Doppler Current Profiler across the Tsugaru Strait (Joint research)

Kawamura, Hideyuki; Hirose, Naoki*; Nakayama, Tomoharu*; Ito, Toshimichi

JAEA-Data/Code 2021-004, 34 Pages, 2021/05

JAEA-Data-Code-2021-004.pdf:3.72MB

The Japan Atomic Energy Agency measured the ocean current across the Tsugaru Strait using an Acoustic Doppler Current Profiler attached on a ferryboat from October 1999 to January 2008. The characteristics of the ocean current in the Tsugaru Strait must be understood for predicting oceanic dispersion of radioactive materials released from nuclear facilities around the strait. Furthermore, it is critical to elucidate the mechanism of the Tsugaru Warm Current from an oceanography viewpoint. The dataset obtained in this investigation consists of daily ocean current data files that record the components of the current speed in the east-west and north-south directions from the surface layer to the bottom layer. The dataset stores 2,211 daily ocean current data files, despite some data periods missing from October 1999 to January 2008. In this study, information on the dataset is described for users to analyze the dataset properly for their purposes. Section 1 provides the background and purpose of the ocean current measurement, Section 2 explains the methodology of measurement using an Acoustic Doppler Current Profiler, and Section 3 explains the record format of the daily ocean current data files and data acquisition rate and presents analysis results. Finally, Section 4 concludes this study.

JAEA Reports

Records of physicochemical parameters by geochemical monitoring system in the Horonobe Underground Research Laboratory (FY2017-FY2019)

Dei, Shuntaro; Mochizuki, Akihito; Miyakawa, Kazuya; Sasamoto, Hiroshi

JAEA-Data/Code 2021-005, 54 Pages, 2021/06

JAEA-Data-Code-2021-005.pdf:4.95MB
JAEA-Data-Code-2021-005-appendix(CD-ROM).zip:5.42MB

Japan Atomic Energy Agency had been conducting "geoscientific study" and "research and development on geological disposal" in the Horonobe Underground Research Laboratory (URL) for safe geological disposal of high-level radioactive waste. Groundwater pressure, pH, and oxidation-reduction potential in the deep groundwater have been continuously monitored with monitoring systems which were developed in the Horonobe URL Project. This report presents the physicochemical parameters of groundwater which have been obtained by the monitoring systems installed at the 140 m, 250 m and 350 m gallery. The data acquired from April 2017 to the end of March 2020 was summarized along with related information such as the specifications of boreholes.

JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.

JAEA Reports

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

JAEA Reports

SCHERN-V2: Technical guide of computer program for chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Data/Code 2021-008, 35 Pages, 2021/08

JAEA-Data-Code-2021-008.pdf:3.68MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. In addition to this, nitrogen oxides (NO$$_{rm x}$$) are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that NOx affects to the migration behavior of Ru at the anticipated atmosphere condition in cells and/or compartments of the facility building. Chemical reactions of NO$$_{rm x}$$ with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. The analysis program, SCHERN has been under developed to simulate chemical behavior including Ru coupled with the thermo-hydraulic condition in the flow paths in the facility building. This technical guide for SCHERN-V2 presents the overview of covered accident, analytical models including newly developed models, differential equations for numerical solution, and user instructions.

JAEA Reports

Update on the regional-scale 3D geological model in the Horonobe Underground Research Laboratory Project

Sakai, Toshihiro; Ishii, Eiichi

JAEA-Data/Code 2021-009, 13 Pages, 2021/08

JAEA-Data-Code-2021-009.pdf:1.9MB
JAEA-Data-Code-2021-009-appendix(CD-ROM).zip:42.79MB

Japan Atomic Energy Agency is performing the Horonobe Underground Research Laboratory Project, which includes a scientific study of the deep geological environment as a basis of research and development for the geological disposal of high level radioactive wastes, in order to establish comprehensive techniques for the investigation, analysis and assessment of the deep geological environment in the sedimentary rock. The numerical data of 3D geological model in regional-scale was compiled in 2019 as JAEA-Data/Code 2019-007, and then this report updates a part of the numerical data of 3D geological model around the underground facilities.

JAEA Reports

Integration of the geological survey data obtained for shaft and gallery walls from the surface to a depth of 380m in the Horonobe Underground Research Laboratory Project

Sakai, Toshihiro; Hayano, Akira

JAEA-Data/Code 2021-010, 243 Pages, 2021/10

JAEA-Data-Code-2021-010.pdf:62.15MB
JAEA-Data-Code-2021-010-appendix1(CD-ROM).zip:95.55MB
JAEA-Data-Code-2021-010-appendix2(CD-ROM).zip:152.69MB
JAEA-Data-Code-2021-010-appendix3(CD-ROM).zip:25.48MB

The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formations at Horonobe, northern Hokkaido. The project consists of two major research areas, "Geoscientific Research" and "R&D on Geological Disposal", and proceeds in three overlapping phases, "Phase I: Surface-based investigation", "Phase II: Construction" and "Phase III: Operation". The geological survey has been carried out at the shafts and the galleries in the Phase II. The geological survey was carried out during the excavation cycle, and the data were obtained for each an excavation cross section. This report shows the data which the individual geological data were integrated for the geological survey at the shafts and the galleries from the surface to a depth of 380m.

JAEA Reports

Analysis of deposits inside the reactor at Fukushima Daiichi Nuclear Power Station in JFY 2017-2018; The Subsidy programs "Project of Decommissioning and Contaminated Water Management in the FY2016 Supplementary Budget, (Development of Technologies for Grasping and Analyzing Properties of Fuel Debris)

Nakayoshi, Akira; Mitsugi, Takeshi; Sasaki, Shinji; Maeda, Koji

JAEA-Data/Code 2021-011, 279 Pages, 2022/03

JAEA-Data-Code-2021-011.pdf:37.76MB

At the TEPCO's Fukushima Daiichi Nuclear Power Station (1F), an investigation inside the reactors has been carried out, and R&D has been made on methods of fuel debris retrieval and storage after retrieval. In order to carry out the decommissioning work safely and steadily, understanding characteristics of fuel debris in the reactors is required. Therefore, in the development of technologies for grasping and analyzing properties of fuel debris project, the characteristics of simulated fuel debris, such as hardness, drying behavior, etc., of fuel debris for design of removal and storage, have been investigated and estimated, and provided to other projects conducting the decommissioning work. As part of this project, U-containing particles in samples (e.g., deposit on the investigation equipment, sediment in the reactors, etc.) obtained during the internal investigation of the reactors of 1F units 1 to 3 were analyzed. This report summarized the results of FE-SEM/WDX, FE-SEM/EDS, STEM/EDS, and TEM analysis, which were extracted from all analysis results obtained, as a database for the evaluation of the generation mechanism of U-containing particles. The analyses were performed at the JAEA Oarai Research and Development Institute and Nippon Nuclear Fuel Development Co., LTD.

JAEA Reports

Development of coupled mass-transport and chemical-reaction calculation code for alteration of engineered barrier

Sasagawa, Tsuyoshi; Mukai, Masayuki; Sawaguchi, Takuma

JAEA-Data/Code 2021-012, 122 Pages, 2022/01

JAEA-Data-Code-2021-012.pdf:3.87MB

Reducing public dose is required when radioactive wastes such as high-level and from reactor core internals etc. are disposed of by means of multi barrier system consist of engineered and natural barriers. In these barriers, engineered barrier is expected to bring out confinement function of waste's radionuclides in the barrier. Materials used as the engineered barriers are altered and performances of the barrier materials are degraded in course of time. To estimate properly the degraded performances, analytical evaluation of long-term change of the engineered barrier state is important. Change state of the engineered barrier is given by mass-transport and geochemical-reaction inside the barrier materials and these phenomena are interrelated, it is necessary to calculate the state by means of coupled analysis procedure. We have developed a coupled mass-transport and geochemical-reaction calculation code (MC- BUFFER) to evaluate alteration of engineered barrier specially targeted for water permeability of bentonite buffer material as one of most important performances to engineered barrier. This report describes functions expected for the engineered barrier, influence parameters for the functions, implementation models in MC-BUFFER, structure and functions of MC-BUFFER, input file format and output examples, execution method of MC-BUFFER, and sample run with MC-BUFFER.

JAEA Reports

Analysis of the radioactivity concentrations in radioactive waste generated from JPDR Facility

Tobita, Minoru*; Haraga, Tomoko; Endo, Tsubasa*; Omori, Hiroyuki*; Mitsukai, Akina; Aono, Ryuji; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2021-013, 30 Pages, 2021/12

JAEA-Data-Code-2021-013.pdf:1.47MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete samples generated from JPDR facility. In this report, we summarized the radioactivity concentrations of 21 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{41}$$Ca, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{rm 108m}$$Ag, $$^{137}$$Cs, $$^{152}$$Eu, $$^{154}$$Eu, $$^{rm 166m}$$Ho, $$^{234}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm) which were obtained from radiochemical analysis of the samples in fiscal year 2018-2019.

JAEA Reports

Analysis system for behavior of water-containing fuel debris; fdradc

Ogawa, Toru

JAEA-Data/Code 2021-014, 139 Pages, 2022/03

JAEA-Data-Code-2021-014.pdf:6.34MB
JAEA-Data-Code-2021-014-hyperlink.zip:5.05MB

A program package has been developed to predict and evaluate the chemical behavior of fuel material mixtures, which are in long-term contact with water. Water radiolysis reaction analysis and electrochemical analysis of the fuel material surface are combined to evaluate the hydrogen evolution and the fuel leaching. The chemical basis of the package modules and their usage are documented.

JAEA Reports

Evaluation of tensile and creep properties on 9Cr-ODS steel claddings

Yano, Yasuhide; Hashidate, Ryuta; Tanno, Takashi; Imagawa, Yuya; Kato, Shoichi; Onizawa, Takashi; Ito, Chikara; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji

JAEA-Data/Code 2021-015, 64 Pages, 2022/01

JAEA-Data-Code-2021-015.pdf:2.6MB

From a view point of practical application of fast breeder reactor cycles, which takes advantage of safety and economic efficiency and makes a contribution of volume reduction and mitigation of degree of harmfulness of high-level radioactive waste, it is necessary to develop fuel cladding materials for fast reactors (FRs) in order to achieve high-burnup. Oxide dispersion strengthened (ODS) steel have been studied for use as potential fuel cladding materials in FRs owing to their excellent resistance to swelling and their high-temperature strength in Japan Atomic Energy Agency. It is very important to establish the materials strength standard in order to apply ODS steels as a fuel cladding. Therefore, it is necessary to acquire the mechanical properties such as tensile, creep rupture strength tests and so on. In this study, tensile and creep rupture strengths of 9Cr-ODS steel claddings were evaluated using by acquired these data. Because of the phase transformation temperature of 9Cr-ODS steel, temperature range for the evaluation was divided into two ones at AC1 transformation temperature of 850$$^{circ}$$C.

JAEA Reports

User manual of NMB4.0

Okamura, Tomohiro*; Nishihara, Kenji; Katano, Ryota; Oizumi, Akito; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

JAEA-Data/Code 2021-016, 43 Pages, 2022/03

JAEA-Data-Code-2021-016.pdf:3.06MB

The quantitative prediction and analysis of the future nuclear energy utilization scenarios are required in order to establish the advanced nuclear fuel cycle. However, the nuclear fuel cycle consists of various processes from front- to back-end, and it is difficult to analyze the scenarios due to the complexity of modeling and the variety of scenarios. Japan Atomic Energy Agency and Tokyo Institute of Technology have jointly developed the NMB code as a tool for integrated analysis of mass balance from natural uranium needs to radionuclide migration of geological disposal. This user manual describes how to create a database and scenario input for the NMB version 4.0.

JAEA Reports

Development of JAEA sorption database (JAEA-SDB); Update of sorption/QA data in FY2021

Sugiura, Yuki; Suyama, Tadahiro*; Tachi, Yukio

JAEA-Data/Code 2021-017, 58 Pages, 2022/03

JAEA-Data-Code-2021-017.pdf:1.98MB

Sorption behavior of radionuclides (RNs) in buffer materials (bentonites), rocks and cementitious materials is one of the key processes in a safe geological disposal of radioactive waste because RNs migration in these materials is expected to be retarded by the sorption process. Therefore, it is necessary to understand the sorption process and develop a database compiling reliable data and mechanistic/predictive models so that reliable parameters can be set under a variety of geochemical conditions relevant to a performance assessment (PA). For this purpose, Japan Atomic Energy Agency (JAEA) has developed the database of sorption parameters in bentonites, rocks and cementitious materials. This sorption database (SDB) was firstly developed as an important basis for the H12 PA of a high-level radioactive waste disposal, and have been provided through the Web. JAEA has continued to improve and update the SDB in the view of potential future needs of data focusing on assuring the desired quality level and testing the usefulness of the databases for possible applications to the PA-related parameter setting. This report focuses on updating of the sorption database (JAEA-SDB) as a basis of integrated approach for the PA-related distribution coefficient (Kd) setting and development of mechanistic sorption models. This report also includes an overview of the database structure and contents. Kd data and their quality assurance (QA) results were updated from literature collected with wider ranges. As a result, 8,503 Kd data from 70 references related to the above-mentioned systems were added and the total number of Kd values in JAEA-SDB reached 79,072. The QA/classified Kd data reached about 75.4% for all Kd data in JAEA-SDB. The updated JAEA-SDB is expected to make it possible to give a basis for the next-step PA-related Kd setting.

JAEA Reports

Calculation code of output current for self-powered radiation detector; Algorithm construction and comparison of calculation results

Shibata, Hiroshi; Takeuchi, Tomoaki; Seki, Misaki; Shibata, Akira; Nakamura, Jinichi; Ide, Hiroshi

JAEA-Data/Code 2021-018, 42 Pages, 2022/03

JAEA-Data-Code-2021-018.pdf:2.78MB
JAEA-Data-Code-2021-018-appendix(CD-ROM).zip:0.15MB

Japan Materials Testing Reactor (JMTR) in Oarai Research and Development Institute of the Japan Atomic Energy Agency has been developing various reactor materials, irradiation techniques and instruments for more than 30 years. Among them, the development of self-powered neutron detectors (SPNDs) and gamma detectors (SPGDs) has been carried out, and several research results have been reported. However, most of the results are based on the design study of the detector development and the results of in-core irradiation tests and gamma irradiation tests using Cobalt-60. In this report, a numerical code is developed based on the paper "Neutron and Gamma-Ray Effects on Self-Powered In-Core Radiation Detectors" written by H.D. Warren and N.H. Shah in 1974, in order to theoretically evaluate the self-powered radiation detectors.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

JAEA Reports

Nuclear Science Research Institute Meteorological Statistics (2006-2020)

Kashimura, Keita; Shoro, Takuya*; Futagawa, Kazuo; Kawasaki, Masatsugu

JAEA-Data/Code 2021-020, 218 Pages, 2022/03

JAEA-Data-Code-2021-020.pdf:2.51MB

These statistical results are based on the meteorological data observed at the Nuclear Science Research Institute in Japan Atomic Energy Agency and statistically processed according to "The guideline of meteorological statistics for the safety analysis of nuclear power reactor" (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The statistics are based on 15 years of meteorological data, from January 2006 to December 2020, processed into five-year periods. These are statistical results of wind direction, wind speed, atmospheric stability, etc., which are used for dose assessment of the general public due to radioactive materials discharged into the atmosphere from nuclear reactor facilities.

JAEA Reports

Data of groundwater chemistry obtained in the Horonobe Underground Research Laboratory Project (FY2021)

Miyakawa, Kazuya

JAEA-Data/Code 2021-021, 23 Pages, 2022/03

JAEA-Data-Code-2021-021.pdf:2.0MB

In the Horonobe underground research laboratory (HURL) project, groundwater chemistry was analyzed to investigate changes due to the excavation of the underground facility and to review geochemical models until the fiscal year 2019. From the fiscal year 2020, to proceed remaining important issues deduced from the conclusion of the investigations during the fiscal year 2015-2019, primary data such as groundwater chemistry need to be successively acquired. Here, the chemical analysis of 54 groundwater samples in 2021 from boreholes drilled in the 140 m, 250 m, 350 m gallery in the HURL, and water rings settled in three vertical shafts is presented. Analytical results include groundwater chemistry such as physicochemical parameters (pH, electrical conductivity), dissolved ions (Na$$^{+}$$, K$$^{+}$$, Li$$^{+}$$, NH$$_{4}$$$$^{+}$$, Cl$$^{-}$$, Br$$^{-}$$, NO$$_{3}$$$$^{-}$$, SO$$_{4}$$$$^{2-}$$, PO$$_{4}$$$$^{3-}$$, Ca$$^{2+}$$, Mg$$^{2+}$$, Sr$$^{2+}$$, P, Total-Mn, Si, Total-Fe, Al, B, F$$^{-}$$, I$$^{-}$$, alkalinity, total organic carbon, total inorganic carbon, CO$$_{3}$$$$^{2-}$$, HCO$$_{3}$$$$^{-}$$, Ba, Fe$$^{2+}$$, sulfide), $$delta$$$$^{18}$$O, $$delta$$D, and tritium content along with a detailed description of analytical methods.

JAEA Reports

Improvement of model for cesium chemisorption onto stainless steel in severe accident analysis code SAMPSON (Joint research)

Miwa, Shuhei; Karasawa, Hidetoshi; Nakajima, Kunihisa; Kino, Chiaki*; Suzuki, Eriko; Imoto, Jumpei

JAEA-Data/Code 2021-022, 32 Pages, 2023/01

JAEA-Data-Code-2021-022.pdf:1.41MB
JAEA-Data-Code-2021-022(errata).pdf:0.17MB

The improved model for cesium (Cs) chemisorption onto stainless steel (SS) in the fission product (FP) chemistry database named ECUME was incorporated into the severe accident (SA) analysis code SAMPSON for the more accurate estimation of Cs distribution within nuclear reactor vessels in the TEPCO's Fukushima Daiichi Nuclear Power Station (1F). The SAMPSON with the improved model was verified based on the analysis results reproducing the experimental results which were subjected to the modeling of Cs chemisorption behavior. Then, the experiment in the facility with the temperature gradient tube to simulate SA conditions such as temperature decrease and aerosol formation was analyzed to confirm availability of the improved model to the analysis of Cs chemisorption onto SS. The SAMPSON with the improved model successfully reproduced the experimental results, which indicates that the improved model and the analytical method such as setting a method of node-junction, models of aerosol formation and the calculation method of saturated CsOH vapor pressure can be applicable to the analysis of Cs chemisorption behavior. As the information on water-solubility of Cs deposits was also prerequisite to estimate the Cs distribution in the 1F because Cs can be transported through aqueous phase after the SA, the water-solubility of chemisorbed Cs compounds was investigated. The chemisorbed compounds on SS304 have been identified to CsFeO$$_{2}$$ at 873 K to 973 K with higher water-solubility, CsFeSiO$$_{4}$$ at 973 K to 1273 K and Cs$$_{2}$$Si$$_{4}$$O$$_{9}$$ at 1073 K to 1273 K with lower water-solubility. From these results, the water-solubility of chemisorbed Cs compounds can be estimated according to the SA analysis conditions such as temperature in the reactor and the CsOH concentration affecting the amount of chemisorbed Cs.

22 (Records 1-22 displayed on this page)
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