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JAEA Reports

Design details of bottom shape for the 3rd glass melter in TVF

Asahi, Yoshimitsu; Shimamura, Keisuke*; Kobayashi, Hidekazu; Kodaka, Akira

JAEA-Technology 2021-026, 50 Pages, 2022/03

JAEA-Technology-2021-026.pdf:6.29MB

In Tokai Reprocessing Plant, the highly active liquid waste derived from a spent fuel reprocessing is vitrified with a Liquid-Fed Ceramic Melter (LFCM) embedded in Tokai Vitrification Facility (TVF). For an LFCM, the viscosity of melted glass is increased by the deposition of oxidation products of platinum group elements (PGE) and the PGE-containing glass tends to settle to the melter's bottom basin even after draining glass out. Removal of the PGE-containing glass is needed to avoid the Joule heating current from being affected by the glass, it requires time-consuming work to remove. For the early accomplishment of vitrifying the waste, Japan Atomic Energy Agency is planning to replace the current melter with the new one in which the amount of PGE sediments would be reduced. In the past design activities for the next melter, several kinds of shapes in regard to the furnace bottom and the strainer were drawn. Among these designs, the one in which the discharge ratio of PGE-containing glass would be as much as or greater than the current melter and which be able to perform similar operational sequences done in the current melter is selected here. Firstly, an operational sequence to produce one canister of vitrified waste is simulated for three melter designs with a furnace bottom shape, using 3D thermal-hydraulic calculations. The computed temperature distribution and its changes are compared among the candidate structures. After discussions about the technical and structural feasibilities of each design, a cone shape with a 45$$^{circ}$$ slope was selected as the bottom shape of the next melter. Secondly, five strainer designs that fit the bottom shape above mentioned are drawn. For each design, the fluid drag and the discharge ratio of relatively high viscosity fluid resting near the bottom are estimated, using steady or unsteady CFD simulation. By draining silicone oil from acrylic furnace models, it was confirmed experimentally that there are no vortices

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

Development of tailor-made adsorbents for uranium recovery from seawater on the basis of uranyl coordination chemistry (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*

JAEA-Review 2021-041, 42 Pages, 2022/01

JAEA-Review-2021-041.pdf:2.03MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Development of tailor-made adsorbents for uranium recovery from seawater on the basis of uranyl coordination chemistry" conducted in FY2020. On the basis of deep understanding on uranyl coordination chemistry, we design molecular structures of pentadentate ligands as functional moieties for uranium adsorption from seawater and study coordination chemistry of uranyl ion with those ligands in order to resolve current problems in uranium recovery technology from seawater and to develop novel selective and efficient adsorbents for this purpose.

Journal Articles

Study of rational safeguards for the treatment of radioactive waste containing nuclear materials

Nakatani, Takayoshi; Shimizu, Ryo; Tazaki, Makiko; Kimura, Takashi; Tamai, Hiroshi; Suda, Kazunori

Dai-42-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2021/11

Currently, JAEA is in the advancing of decommissioning own nuclear facilities, and some facilities have been handling nuclear materials. In decommissioning, it is necessary to consider rational methods while keeping nuclear non-proliferation and transparency, including treatment of radioactive waste generated from these facilities and methods for terminating safeguards. In this study, we considered the above issues regarding waste treatment with reference to the guidance of Safeguards by Design (SBD) published by the International Atomic Energy Agency (IAEA).

JAEA Reports

Report of summer holiday practical training 2020; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 3

Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (18$$times$$3 layers) fuel blocks with 20% enrichment of $$^{235}$$U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.

Journal Articles

Great achievements of M. Salvatores for nuclear data adjustment study with use of integral experiments

Yokoyama, Kenji; Ishikawa, Makoto*

Annals of Nuclear Energy, 154, p.108100_1 - 108100_11, 2021/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the design of innovative nuclear reactors such as fast reactors, the improvement of the prediction accuracies for neutronics properties is an important task. The nuclear data adjustment is a promising methodology for this issue. The idea of the nuclear data adjustment was first proposed in 1964. Toward its practical application, however, a great deal of study has been conducted over a long time. While it took about 10 years to establish the theoretical formulation, the research and development for its practical application has been conducted for more than half a century. Researches in this field are still active, and the fact suggests that the improvement of the prediction accuracies is indispensable for the development of new types of nuclear reactors. Massimo Salvatores, who passed away in March 2020, was one of the first proposers to develop the nuclear data adjustment technique, as well as one of the great contributors to its practical application. Reviewing his long-time works in this area is almost the same as reviewing the history of the nuclear data adjustment methodology. The authors intend that this review would suggest what should be done in the future toward the next development in this area. The present review consists of two parts: a) the establishment of the nuclear data adjustment methodology and b) the achievements related to practical applications. Furthermore, the former is divided into two aspects: the study on the nuclear data adjustment theory and the numerical solution for sensitivity coefficient that is requisite for the nuclear data adjustment. The latter is separated to three categories: the use of integral experimental data, the uncertainty quantification and design target accuracy evaluation, and the promotion of nuclear data covariance development.

JAEA Reports

Investigation and consideration on evaluation of radiation doses to residents in the case of a nuclear emergency

Hashimoto, Makoto; Kinase, Sakae; Munakata, Masahiro; Murayama, Takashi; Takahashi, Masa; Takada, Chie; Okamoto, Akiko; Hayakawa, Tsuyoshi; Sukegawa, Masato; Kume, Nobuhide*; et al.

JAEA-Review 2020-071, 53 Pages, 2021/03

JAEA-Review-2020-071.pdf:2.72MB

In the case of a nuclear accident or a radiological emergency, the Japan Atomic Energy Agency (JAEA), as a designated public corporation assigned in the Disaster Countermeasures Basic Act and the Armed Attack Situation Response Law, undertakes technical supports to the national government and local governments. The JAEA is requested to support to evaluate radiation doses to residents in a nuclear emergency, which is specified in the Basic Disaster Management Plan and the Nuclear Emergency Response Manual. For the dose evaluation, however, its strategy, target, method, structure and so on have not been determined either specifically or in detail. This report describes the results of investigation and consideration discussed in the "Working Group for Radiation Dose Evaluation at a Nuclear Emergency" established within the Nuclear Emergency Assistance and Training Center to discuss technical supports for radiation dose evaluation to residents in the case of a nuclear emergency, and aims at contributing to specific and detailed discussion and activities in the future for the national government and local governments, also within the JAEA.

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Journal Articles

Derivation of ideal power distribution to minimize the maximum kernel migration rate for nuclear design of pin-in-block type HTGR

Okita, Shoichiro; Fukaya, Yuji; Goto, Minoru

Journal of Nuclear Science and Technology, 58(1), p.9 - 16, 2021/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Suppressing the kernel migration rates, which depend on both the fuel temperature and the fuel temperature gradient, under normal operation condition is quite important from the viewpoint of the fuel integrity for High Temperature Gas-cooled Reactors. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.

Journal Articles

Criticality configuration design methodology applied to the design of fuel debris experiment in the new STACY

Gunji, Satoshi; Tonoike, Kotaro; Clavel, J.-B.*; Duhamel, I.*

Journal of Nuclear Science and Technology, 58(1), p.51 - 61, 2021/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k$$_{rm eff}$$ sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k$$_{rm eff}$$ sensitivity to the cross section. For this purpose, some parameters of the core configuration, as for example the lattice pitches or the core dimensions, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.

Journal Articles

Issues and recommendations about application of graded approach to research reactors

Yonomoto, Taisuke; Mineo, Hideaki; Murayama, Yoji; Hohara, Shinya*; Nakajima, Ken*; Nakatsuka, Toru; Uesaka, Mitsuru*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(1), p.73 - 77, 2021/01

no abstracts in English

JAEA Reports

Development of tailor-made adsorbents for uranium recovery from seawater on the basis of uranyl coordination chemistry (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*

JAEA-Review 2020-026, 41 Pages, 2020/12

JAEA-Review-2020-026.pdf:3.25MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2019. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Development of Tailor-Made Adsorbents for Uranium Recovery from Seawater on the Basis of Uranyl Coordination Chemistry". On the basis of deep understanding on uranyl coordination chemistry, we design molecular structures of pentadentate ligands as functional moieties for uranium adsorption from seawater and study coordination chemistry of uranyl ion with those ligands in order to resolve current problems in uranium recovery technology from seawater and to develop novel selective and efficient adsorbents for this purpose.

Journal Articles

Voltage drop analysis and leakage suppression design for mineral-insulated cables

Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Otsuka, Noriaki; Tsuchiya, Kunihiko

Journal of Nuclear Science and Technology, 57(12), p.1276 - 1286, 2020/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The influence of materials of mineral-insulated (MI) cables on their electrical characteristics upon exposure to high-temperature conditions was examined via a transmission test, in the objective of achieving the stability of the potential distribution along the cable length. Occurrence of a voltage drop along the cable was confirmed for aluminum oxide (Al$$_{2}$$O$$_{3}$$) and magnesium oxide (MgO), as insulating materials of the MI cable. A finite-element method (FEM)-based analysis was performed to evaluate the leakage in the potentials, which was found at the terminal end. Voltage drop yields by the transmission test and the analysis were in good agreement for the MI cable of Al$$_{2}$$O$$_{3}$$ and MgO materials, which suggests the reproducibility of the magnitude relationship of the experimental results via the FEM analysis. To suppress the voltage drop, the same FEM analysis was conducted, the diameter of the core wires ($$d$$) and the distance between them ($$l$$) were varied. Considering the variation of $$d$$, the potential distribution in the MI cable produced a minimum voltage drop corresponding to a ratio $$d/D$$ of 0.35, obtained by dividing $$d$$ with that of the insulating material ($$D$$). In case of varying $$l$$, a minimum voltage drop was l/$$D$$ of 0.5.

Journal Articles

Pressure resistance thickness of disposal containers for spent fuel direct disposal

Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09

A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.

Journal Articles

Neutronic design of neutron moderator on a reentrant-hole configuration for Kyoto University Accelerator-based Neutron Source (KUANS)

Okita, Shoichiro; Tasaki, Seiji*; Abe, Yutaka*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.178 - 184, 2020/09

The Kyoto University Accelerator-based Neutron Source (KUANS) is a compact neutron source that is mainly used for spectrometer and detector development. In addition, it is also suited for experiments to study the neutronic design of moderators owing to the relatively low neutron generation yield by $$^{9}$$Be(p,n). We present a neutronic design of the neutron moderator on a reentrant-hole configuration for KUANS to enhance the neutron emission, and some experiments are conducted at KUANS for verification. A polyethylene moderator on a reentrant-hole configuration is designed by PHITS calculation and is introduced to KUANS to obtain intense oblong neutron beams. The intensity of the pulsed neutron beam is experimentally measured. The results reveal that the intensity becomes approximately 1.9 times stronger than that of the conventional rectangular design. In addition, the ratio of its intensity to the conventional intensity increases to approximately threefold as the neutron wavelength increases. It is interesting to note that the longer the neutron wavelength, the more efficiently they are extracted from the inside of the moderator owing to the existence of the reentrant-hole configuration.

JAEA Reports

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the $$^{235}$$U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with $$^{235}$$U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.

Mechanical Engineering Journal (Internet), 7(3), p.19-00489_1 - 19-00489_16, 2020/06

The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.

JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements (Translated document)

Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru

JAEA-Technology 2019-020, 167 Pages, 2020/03

JAEA-Technology-2019-020.pdf:21.06MB
JAEA-Technology-2019-020-high-resolution1.pdf:47.3MB
JAEA-Technology-2019-020-high-resolution2.pdf:34.99MB
JAEA-Technology-2019-020-high-resolution3.pdf:48.74MB
JAEA-Technology-2019-020-high-resolution4.pdf:47.83MB
JAEA-Technology-2019-020-high-resolution5.pdf:18.35MB
JAEA-Technology-2019-020-high-resolution6.pdf:49.4MB
JAEA-Technology-2019-020-high-resolution7.pdf:39.78MB

The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.

Journal Articles

Effects of heterogeneity of geomechanical properties on tunnel support stress during tunnel excavation

Okazaki, Yasuyuki*; Hayashi, Hisashi*; Aoyagi, Kazuhei; Morimoto, Shingo*; Shinji, Masato*

Proceedings of 5th ISRM Young Scholars' Symposium on Rock Mechanics and International Symposium on Rock Engineering for Innovative Future (YSRM 2019 and REIF 2019) (USB Flash Drive), 6 Pages, 2019/12

In the design of tunnel support, the behavior of the rock mass around a tunnel and the stress acting on the tunnel support may be predicted using a numerical analysis. However, in such a numerical analysis, it is common to assume that each stratum comprises a homogeneous material, ignoring the heterogeneity of the geomechanical properties inherent to the rock mass. For this reason, it is not unusual for the results of the numerical analysis to differ from the actual behavior. We performed a tunnel excavation analysis considering the heterogeneity of the geomechanical properties in the rock mass to investigate the local increase in the tunnel support stress obtained in the 350 m gallery at the Horonobe Underground Research Laboratory. The results revealed that, in order to predict the locally increased support stress in advance, it is necessary to carry out a tunneling excavation analysis considering the heterogeneity of the geomechanical properties. It was also revealed that the scale at which the geomechanical properties fluctuate is an important factor.

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