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Journal Articles

Development of nuclear data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

Journal of Nuclear Science and Technology, 61(6), p.830 - 839, 2024/06

 Times Cited Count:3 Percentile:75.12(Nuclear Science & Technology)

Nuclear data processing code is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, consideration of the resonance upscattering, ACE file perturbation, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.

Journal Articles

Investigation of the impact of difference between FRENDY and NJOY2016 on neutronics calculations

Ono, Michitaka*; Tojo, Masayuki*; Tada, Kenichi; Yamamoto, Akio*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05

In this paper, nuclear calculations were performed using the ACE files and the multigroup libraries created by both FRENDY and NJOY, and the impacts on the neutronics characteristics due to nuclear data processing were investigated using those libraries. MCNP was used to compare the ACE files by calculating many benchmark problems including ICSBEP and it was confirmed that the k-eff values are generally agreed with each other within the range of statistical errors. The multigroup cross sections are verified by the BWR design codes LANCR/AETNA through calculation of a commercial-grade BWR5 equilibrium core loaded with 9$$times$$9 fuels. These results indicate that fuel assembly and core characteristics are consistent with each other. From the above investigations, it was confirmed that FRENDY can provide comparable continuous/multi-group neutron cross sections with NJOY.

Journal Articles

Development of nuclear data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

Nuclear data processing is an important interface between an evaluated nuclear data library and nuclear transport calculation codes. JAEA has developed a new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates ACE files which are used for the continuous-energy Monte Carlo codes including PHITS, Serpent, and MCNP; it was released as an open-source software under the 2-clause BSD license in 2019. After FRENDY version 1 was released, many functions are developed: the multi-group neutron cross-section library generation, the statistical uncertainty quantification for the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, etc. We released FRENDY version 2 including these functions. This presentation explains the overview of FRENDY and features of the new functions implemented in FRENDY version 2.

Journal Articles

Recent activities in the field of reactor physics

Fukushima, Masahiro; Tojo, Masayuki*

Journal of Nuclear Science and Technology, 56(12), p.1061 - 1062, 2019/12

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Reactor Physics that treat the essentials of how fission nuclear reactors work fundamentally has important roles on safe operations and design studies of various types of nuclear reactors. From the latest activities in the field of reactor physics, this report summarizes some outstanding researches and developments published in scientific journals including the Journal of Nuclear Science and Technology.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 2; Fuel cladding oxidation

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

Oxidation behaviour of Zr cladding in SFP accident condition was evaluated by using a thermobalance in this work, and the obtained data were applied to construct oxidation model for SFP accident condition. For the validation of the constructed oxidation model, oxidation tests using a long cladding tube 500mm in length were conducted in conditions simulating SFP accidents, such as flow rate of the atmosphere in spent fuel rack, temperature gradient along the axis of cladding, and heating-up history. Thickness of oxide layer formed on the surface of cladding samples was evaluated by cross sectional observation, and compared with calculation results obtained by using the oxidation model. The detail of experimental results and validation of the oxidation model will be discussed.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 4; Investigation of fuel loading effects in BWR spent fuel rack

Tojo, Masayuki*; Kanazawa, Toru*; Nakashima, Kazuo*; Iwamoto, Tatsuya*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In this study, fuel loading effects in BWR spent fuel rack accidents are widely investigated using three-dimensional analysis methods from both nuclear and thermal hydraulics viewpoints, including: (a) Decay heat of spent fuel after discharge, (b) The maximum temperature of spent fuel cladding in the spent fuel rack depending on heat transfer phenomena, and (c) Criticality of the spent fuel rack after collapsing of the fuel due to a severe accidents in the BWR spent fuel pool (SFP).

Journal Articles

Study on oxidation model for Zircalloy-2 cladding in SFP accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Onizawa, Takashi*; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2018) (USB Flash Drive), 8 Pages, 2018/09

The authors proposed oxidation models based on oxidation data which previously obtained in high temperature oxidation tests on small sample of Zircalloy-2 (Zry2) cladding in dry air and in air/steam mixture environment. The oxidation models were implemented in computational fluid dynamics (CFD) code to analyse oxidation behavior of long cladding sample in hypothetical spent fuel pool (SFP) accident conditions. The oxidation tests were conducted using Zry2 cladding sample 500 mm in length. The oxide layer growth in dry air was well reproduced in the calculation using the oxidation model, meanwhile which in air/steam mixture was overestimated atmosphere composition change anticipated in the spent fuel rack during the accident, and its influence on the oxidation behaviour of the cladding were discussed in consideration of the oxidation model improvement.

Journal Articles

Influence of the air/steam mixing ratio in atmosphere on zirconium cladding oxidation in spent fuel pool accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

To cope with the hypothetical severe accident in spent fuel pools (SFPs), it is important to understand the high temperature oxidation behavior of the Zirconium claddings exposed in the air or in the atmosphere of air/steam mixture. In this study, oxidation tests on Zircaloy-2 (Zry2) and Zircaloy-4 (Zry4) short samples were conducted in the atmosphere of air - steam mixture, and mixing ratio was varied to evaluate its influence on the oxidation kinetics in the temperature range from 600 to 1100$$^{circ}$$C. From 900 to 1000$$^{circ}$$C for Zry2, and from 800 to 1000$$^{circ}$$C for Zry4, oxidation rates appeared higher in air - steam mixture than in dry air or in steam without air. This tendency was appeared more evident in post-breakaway transition phase after fracture of dense oxide layer on the surface of specimens. These results suggest importance of the oxidation model development in consideration of the air - steam mixture environment for the SFP accident analysis.

Journal Articles

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo*; Kanazawa, Toru*; Tojo, Masayuki*

Journal of Nuclear Materials, 488, p.22 - 32, 2017/05

AA2016-0383.pdf:0.86MB

 Times Cited Count:3 Percentile:18.14(Materials Science, Multidisciplinary)

The authors previously conducted thermogravimetric analyses on zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.

Journal Articles

Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo*; Tojo, Masayuki*

Zairyo To Kankyo, 66(5), p.180 - 187, 2017/05

In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures and different air flow rates in this work. Oxidation rate increased with temperature. In range of air flow rate predicted in spent fuel lack during SFP accident, influence of flow rate was not clearly observed below 950$$^{circ}$$C for Zry2 and below 1050$$^{circ}$$C for Zry4. Over these temperature, oxidation rates appeared obviously higher in higher air flow rate, and this trend became clearer when temperature increased. Oxide layers were carefully examined after the oxidation tests and compared with the mass gain data in TGA to investigate detail of air oxidation process. The results revealed that mass gain data in the pre breakaway transition stage reflects growth of the dense oxide film on specimen surface, and in the post breakaway transition stage, it reflects growth of porous oxide layer beneath the breakaway cracking of the oxide film.

Journal Articles

Study on oxidation behavior of cladding for accident conditions in spent fuel pool

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo*; Tojo, Masayuki*; Goto, Daisuke*

Fushoku Boshoku Kyokai Dai-62-Kai Zairyo To Kankyo Toronkai Koenshu (CD-ROM), p.23 - 24, 2015/11

In order to clarify the air oxidation behavior of the cladding at high temperatures for study on improvement of safety for accident conditions in spent fuel pool, the oxidation tests for both small specimens under constant temperature conditions and long specimens under loss of coolant simulated temperature conditions were carried out, and the knowledge for influence of both temperature gradient and preoxide film on oxidation behavior of the cladding were obtained in this study.

Oral presentation

Investigation of the impact of difference between open nuclear data processing codes on neutron transport calculations, 2; Investigation of the impact of difference between nuclear data processes on nuclear calculations

Ono, Michitaka*; Tojo, Masayuki*; Tada, Kenichi; Ikehara, Tadashi

no journal, , 

The difference of the nuclear data processing affects the difference of cross section library and neutronics calculation. This presentation explains the comparisons of the cross section libraries and neutronics calculation results.

Oral presentation

Investigation of the impact of difference between open nuclear data processing codes on neutron transport calculations, 5; Investigation of the impact of difference between multigroup nuclear data processes on nuclear calculations

Ono, Michitaka*; Tojo, Masayuki*; Yamamoto, Akio*; Tada, Kenichi

no journal, , 

FRENDY-MG enables us to generate multi-group XS library. We investigated the impact of difference between multi-group nuclear data processes on nuclear calculations.

Oral presentation

Study on improvement of safety for accident conditions in spent fuel pool, 21; Investigation of cladding oxidation model

Nemoto, Yoshiyuki; Onizawa, Takashi*; Kaji, Yoshiyuki; Nakashima, Kazuo*; Kanazawa, Toru*; Tojo, Masayuki*

no journal, , 

Cladding oxidation model construction and its validation is necessary for the upgrading of severe accidentcode which will be utilized for analysis of the spent fuel pool (SFP) accident. Cladding oxidation tests and analyses using the proposed oxidation model were coducted in the simulated SFP accident condition, and their results were compared. Modification of the oxidation model will be discussed in the lecture.

Oral presentation

Study on improvement of safety for accident conditions in spent fuel pool, 23; CFD analysis for the loss of coolant accident in SFP with Zircalloy-2 cladding oxidation model

Tojo, Masayuki*; Kanazawa, Toru*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

no journal, , 

Computational fluid dynamics (CFD) analysis for the loss of coolant accident in spent fuel pool (SFP) with Zircalloy-2 cladding was conducted. In the higher region of spent fuel rack internal, oxygen content in the atmosphere was decreased because of oxigen consumption during the oxidation reaction of cladding material. This oxygen depression is thought to surpress the oxidation of claddings.

Oral presentation

Study on improvement of safety for accident conditions in spent fuel pool, 24; Create a data base of BWR fuel temperature in SFP based on the steady-state heat transfer analyses

Goto, Daisuke*; Kobayashi, Kensuke*; Tojo, Masayuki*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

no journal, , 

Influence of base temperature of the spent fuel pool (SFP), rack dimension, structure, and decay heat distribution, for the maximum temperature of cladding in spent fuel rack during the SFP loss of coolant accident was evaluated in this study, by using the previously developed 3 dimensional steady-state heat transfer analysis method.

Oral presentation

Study on improvement of safety for accident conditions in spent fuel pool, 25; Experiment and analysis of the water level mismatch between in and out of CB of BWR fuel under Loss of Cooling System

Kobayashi, Kensuke*; Goto, Daisuke*; Tojo, Masayuki*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

no journal, , 

Experiment and analysis simulating the water level mismatch between inside and outside of chanel box of BWR fuel under loss of cooling system were conducted for validation of the equation for void fraction correlation.

Oral presentation

Nuclear data processing code FRENDY Version 2

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

no journal, , 

FRENDY (From Evaluated Nuclear Data library to any application) is a nuclear data processing code for the evaluated nuclear data libraries JENDL, ENDF/B, JEFF, TENDL, and so on. The first version of FRENDY was released in 2019 as an open-source software under the 2-clause BSD license. FRENDY Ver. 1 generates ACE files which is used for the continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. Many new modules, e.g., the multi-group neutron cross-section generation from the ACE file, are implemented in FRENDY Ver. 2. This presentation explains the characteristics of FRENDY and new capabilities implemented in FRENDY Ver. 2.

Oral presentation

Development of nuclear data processing code FRENDY version 2, 1; Overview of FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

no journal, , 

JAEA released the nuclear data processing code FRENDY version 2 in January 2022. This presentation explains the overview of new functions implemented in FRENDY version 2, e.g., multi-group cross section generation, uncertainty quantification for probability tables, perturbation of ACE file, and modification of evaluated nuclear data file.

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