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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

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

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.

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

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.

JAEA Reports

Safety design report on J-PARC Transmutation Physics Experimental Facility (TEF-P)

Partitioning and Transmutation Technology Division, Nuclear Science and Engineering Center

JAEA-Technology 2017-033, 383 Pages, 2018/02

JAEA-Technology-2017-033.pdf:28.16MB

JAEA is pursuing research and development (R&D) on volume reduction and mitigation of degree of harmfulness of high-level radioactive waste. Construction of Transmutation Experimental Facility (TEF) is under planning as one of the second phase facilities in the Japan Proton Accelerator Complex (J-PARC) program to promote R&D on the transmutation technology with using accelerator driven systems (ADS). The TEF consists of two facilities: ADS Target Test Facility (TEF-T) and Transmutation Physics Experimental Facility (TEF-P). Development of spallation target technology and study on target materials are to be conducted in TEF-T with impinging a high intensity proton beam on a liquid lead-bismuth eutectic target. Whereas in TEF-P, by introducing a proton beam to minor actinide loaded cores, reactor physical properties of the cores are to be studied, and operation experiences of ADS are to be acquired. This report summarizes results of safety design for establishment permit of one of two TEF facilities, TEF-P.

JAEA Reports

Development of fuel temperature calculation code "FTCC" for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju

JAEA-Data/Code 2017-002, 74 Pages, 2017/03

JAEA-Data-Code-2017-002.pdf:2.36MB

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

Journal Articles

Development of fuel temperature calculation code for HTGRs

Inaba, Yoshitomo; Nishihara, Tetsuo

Annals of Nuclear Energy, 101, p.383 - 389, 2017/03

 Times Cited Count:6 Percentile:65.17(Nuclear Science & Technology)

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.

Journal Articles

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

 Times Cited Count:1 Percentile:14.27(Nuclear Science & Technology)

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.

Journal Articles

Improvement of transient analysis method of a sodium-cooled fast reactor with FAIDUS fuel sub-assemblies

Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code $$alpha$$-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.

JAEA Reports

Development of fuel temperature calculation file for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Fukaya, Yuji; Tachibana, Yukio

JAEA-Data/Code 2014-023, 64 Pages, 2015/01

JAEA-Data-Code-2014-023.pdf:7.15MB

The Japan Atomic Energy Agency has performed the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the deployment of the systems to overseas such as developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. In the core thermal and hydraulic designs of HTGRs, it is important to evaluate the maximum fuel temperature so that the thermal integrity of the fuel is ensured. In order to calculate and evaluate the fuel temperature on personal computers (PCs) in a convenient manner, the calculation file based on the Microsoft Excel were developed. In this report, the basic equations used in the calculation file, the calculation method and procedure, and the results of the validation calculation are described.

JAEA Reports

Report on the 8th Workshop on the Innovative Water Reactor for Flexible Fuel Cycle; February 10, 2005, Koku-kaikan, Minato-ku, Tokyo

Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

JAERI-Review 2005-029, 119 Pages, 2005/09

JAERI-Review-2005-029.pdf:11.01MB

The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.

Journal Articles

Core thermal-hydraulic design

Takada, Eiji*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke

Nuclear Engineering and Design, 233(1-3), p.37 - 43, 2004/10

 Times Cited Count:10 Percentile:58.07(Nuclear Science & Technology)

The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88 % of the total flow is achieved at minimum. The maximum fuel temperature appears during the high temperature test operation, and reaches 1492 $$^{circ}$$C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 $$^{circ}$$C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 $$^{circ}$$C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 $$^{circ}$$C. It is confirmed that the core thermal-hydraulic design gives conservative results.

Journal Articles

Introduction to modern nodal method and discontinuity factor

Okumura, Keisuke

Nihon Genshiryoku Gakkai Dai-36-Kai Robutsuri Kaki Semina Tekisuto, p.81 - 102, 2004/08

The modern node method which uses a discontinuous factor has come to be widely used recently in the reactor core analyses of commercial light water reactors. The basic theory, numerical computation technique and examples of calculation results are explained for biginners of the modern nodal method.

Journal Articles

Nuclear, thermal and hydraulic design for Gas Turbine High Temperature Reactor (GTHTR300)

Nakata, Tetsuo*; Katanishi, Shoji; Takada, Shoji; Yan, X.; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(4), p.478 - 489, 2003/12

no abstracts in English

JAEA Reports

Summary of the 6th Workshop on the Reduced-Moderation Water Reactor; March 6, 2003, JAERI, Tokai

Nabeshima, Kunihiko; Nakatsuka, Toru; Ishikawa, Nobuyuki; Uchikawa, Sadao

JAERI-Conf 2003-020, 240 Pages, 2003/11

JAERI-Conf-2003-020.pdf:27.66MB

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since 1998 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The workshop began with five lectures on status of research and development on RMWRs in JAERI entitled "Status and Future Program of Research and Development on Reduced-Moderation Water Reactors", "Design of Small Reduced-Moderation Water Reactors", "Critical Experiments for Reduced-Moderation Water Reactors", "Critical Heat Flux Experiments in Tight Lattice Core" and "Development of High Performance Cladding". Then two lectures followed: "Status of Phase II of Feasibility Studies on Commercialized Fast Breeder Reactor System" by JNC and "Present Status of Study on Super-critical water Cooled Power Reactor" by Toshiba Corporation.

Journal Articles

Development of Reduced-Moderation Water Reactor (RMWR) for sustainable energy supply

Iwamura, Takamichi; Okubo, Tsutomu; Kureta, Masatoshi; Nakatsuka, Toru; Takeda, Renzo*; Yamamoto, Kazuhiko*

Proceedings of 13th Pacific Basin Nuclear Conference (PBNC 2002) (CD-ROM), 7 Pages, 2002/10

In order to ensure sustainable energy supply in Japan, the reduced-moderation water reactor (RMWR) has been developed by JAERI since 1998. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio. In order to establish negative void reactivity coefficient, the core should be short and flat to increase neutron leakage from the core. The core designs were accomplished to a large core with 1,356MWe and a small core with 330MWe. For both cores, negative void coefficient and natural circulation cooling of the core were realized. To confirm thermal-hydraulic feasibility, critical heat flux experiments were performed using 7-rod bundles with the gap width of 1mm and 1.3mm. The results indicated that enough cooling was assured for the tight lattice core. Further R&D studies, including large scale thermal-hydraulic experiments, reactor physics experiments, development of high burn-up fuel cladding material and simplified reprocessing technology, are necessary to realize commercial introduction of RMWR by 2020's for the replacement of current generation LWRs.

Journal Articles

RAPID-L highly automated fast reactor concept without any control rods, 2; Critical experiment of lithium-6 used in LEM and LIM

Tsunoda, Hirokazu*; Sato, Osamu*; Okajima, Shigeaki; Yamane, Tsuyoshi; Iijima, Susumu; Kobe, Mitsuru*

Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 6 Pages, 2002/00

no abstracts in English

JAEA Reports

Research on reduced-moderation water reactor (RMWR)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro*; Usui, Shuji*; Shirakawa, Toshihisa*; Nakatsuka, Toru; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Wada, Shigeyuki*

JAERI-Research 99-058, p.61 - 0, 1999/11

JAERI-Research-99-058.pdf:3.3MB

no abstracts in English

JAEA Reports

MOSRA-Light; High speed three-dimensional nodal diffusion code for vector computers

Okumura, Keisuke

JAERI-Data/Code 98-025, 243 Pages, 1998/10

JAERI-Data-Code-98-025.pdf:10.15MB

no abstracts in English

83 (Records 1-20 displayed on this page)