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Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

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

https://doi.org/10.1016/B978-0-12-803581-8.11682-0

Journal Articles

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

Miyahara, Naoya; Miwa, Shuhei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko

Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02

https://doi.org/10.1080/00223131.2018.1544939

Times Cited Count：7 Percentile：61.94(Nuclear Science & Technology)

In order to improve LWR source term under severe accident conditions, the first version of a fission product (FP) chemistry database named "ECUME" was developed. The ECUME is intended to include major chemical reactions and their effective kinetic constants for representative SA sequences. It is expected that the ECUME can serve as a fundamental basis from which FP chemical models in the SA analysis codes can be elaborated. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate FP chemistry in Cs-I-B-Mo-O-H system under LWA SA conditions.

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.

JAEA Reports

Application of FORNAX-A

Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo

JAEA-Technology 2015-040, 32 Pages, 2016/02

JAEA-Technology-2015-040.pdf:0.83MB

Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.

JAEA Reports

HTFP for calculation of amount of additionally released fission products from fuel rods of pin-in-block-type high temperature gas-cooled reactors during accident

Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi

JAEA-Data/Code 2015-008, 39 Pages, 2015/06

JAEA-Data-Code-2015-008.pdf:10.32MB

HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.

Journal Articles

Parametric survey on possible impact of partitioning and transmutation of high-level radioactive waste

Oigawa, Hiroyuki; Yokoo, Takeshi*; Nishihara, Kenji; Morita, Yasuji; Ikeda, Takao*; Takaki, Naoyuki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The benefit of implementing Partitioning and Transmutation (P&T) of high-level wastes was parametrically surveyed. The possible reduction of the geological repository area was estimated. By recycling minor actinides (MA), the repository area required for unit spent fuel was reduced significantly in the case of MOX-LWR. This effect was caused by removal of $^{241}$ Am which is a long-term heat source. By partitioning the fission products, in addition to MA recycling, further 70-80% reduction from the MA-recovery case can be expected for both UO and MOX. This significant reduction was independent of the cooling time before the partitioning process.

JAEA Reports

Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

Sugiyama, Tomoyuki; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Nagase, Fumihisa; Fuketa, Toyoshi

JAERI-Research 2003-033, 76 Pages, 2004/01

JAERI-Research-2003-033.pdf:17.46MB

Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 88BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. Fission gas release during the pulse irradiation was from 3.1% to 8.2%, depending on the peak fuel enthalpy and the normal operation conditions.

Journal Articles

Validation of minor actinide cross sections by studying samples irradiated for 492 days at the dounreay prototype fast reactor, 2; Burnup calculations

Tsujimoto, Kazufumi; Kono, Nobuaki; Shinohara, Nobuo; Sakurai, Takeshi; Nakahara, Yoshinori; Mukaiyama, Takehiko; Raman, S.*

Nuclear Science and Engineering, 144(2), p.129 - 141, 2003/06

https://doi.org/10.13182/NSE03-A2348

To evaluate neutron cross-section data of minor actinides, separated actinide samples and dosimetry samples were irradiated at the Dounreay Prototype Fast Reactor for 492 effective full power days. Based on the burnup calculations of major actinide and dosimetry samples, the neutron flux distribution and the flux level were adjusted at the locations where minor actinide samples were irradiated. The burnup calculations were carried out for minor actinides using the determined flux distribution and flux level. This paper discusses the burnup calculations and the validation of minor actinide cross-section data in evaluated nuclear data libraries. We find that we can obtain reliable FIMA (fission per initial metallic atom) values by using the $^{148}$ Nd method except that the uncertainties in the FIMA values are large for $^{234}$ U, $^{238}$ Pu, Am isotopes, and Cm isotopes because the $^{148}$ Nd yields are known poorly for these isotopes and are probably overestimated. For these isotopes, measurements to improve the fission-yield data are needed. We also find that, in general, the JENDL-3.2 nuclear data for the minor actinides are adequate for the conceptual design study of transmutation systems. But, there are some nuclides (especially $^{238}$ Pu and $^{242}$ Pu) for which new measurements are needed particulary if the minor actinides constitute a major part of the nuclear fuel.

JAEA Reports

Descriptive data of JENDL-3.3

Shibata, Keiichi

JAERI-Data/Code 2002-026, 516 Pages, 2003/01

JAERI-Data-Code-2002-026-Part1.pdf:25.12MB
JAERI-Data-Code-2002-026-Part2.pdf:28.24MB

The third revision of JENDL-3 (JENDL-3.3) was released in 2002. The library contains evaluated neutron nuclear data for 337 nuclides. This report presents a brief description of the evaluation method which is given in the MF1 part of JENDL-3.3.

Journal Articles

Conceptual design study of the transmutation experimental facilities

Sasa, Toshinobu; Oigawa, Hiroyuki; Kikuchi, Kenji; Ikeda, Yujiro

Proceedings of American Nuclear Society Conference "Nuclear Applications in the New Millennium" (AccApp-ADTTA '01) (CD-ROM), 7 Pages, 2002/00

A design study of the accelerator-driven system (ADS) for transmutation of minor actinides and long-lived fission products has been performed at Japan Atomic Energy Research Institute (JAERI) under the national OMEGA program. To solve the technical issues related to the ADS development, a transmutation experimental facility (TEF) is planned to bulid under the JAERI-High Energy Accelerator Research Organization joint project. The TEF consists of two facilities, Transmutation Physics Experimental Facility and Transmutation Engineering Experimental Facility. Proton beams of 600 MeV and 0.3 mA are to be delivered to this facility. The presentation describes a conceptual design study of the TEF.

Journal Articles