Conceptual study of Post Irradiation Examination (PIE) Facility at J-PARC

Saito, Shigeru; Meigo, Shinichiro; Makimura, Shunsuke*; Hirano, Yukinori*; Tsutsumi, Kazuyoshi*; Maekawa, Fujio

JAEA-Technology 2023-025, 48 Pages, 2024/03

JAEA-Technology-2023-025.pdf:3.11MB

JAEA has been developing Accelerator-Driven Systems (ADS) for research and development of nuclear transmutation using accelerators in order to reduce the volume and hazardousness of high-level radioactive waste generated by nuclear power plants. In order to prepare the material irradiation database necessary for the design of ADS and to study the irradiation effects in Lead-Bismuth Eutectic (LBE) alloys, a proton irradiation facility is under consideration at J-PARC. In this proton irradiation facility, 250 kW proton beams will be injected into the LBE spallation target, and irradiation tests under LBE flow will be performed for candidate structural materials for ADS. Furthermore, semiconductor soft-error tests, medical RI production, and proton beam applications will be performed. Among these, Post Irradiation Examination (PIE) of irradiated samples and RI separation and purification will be carried out in the PIE facility to be constructed near the proton irradiation facility. In this PIE facility, PIE of the equipment and samples irradiated in other facilities in J-PARC will also be performed. This report describes the conceptual study of the PIE facility, including the items to be tested, the test flow, the facilities, the test equipment, etc., and the proposed layout of the facility.

Development and validation of analysis code for spallation products behavior in LBE coolant system of ADS comparing with the distribution data in MEGAPIE spallation target

Miyahara, Shinya*; Arita, Yuji*; Nakano, Keita; Maekawa, Fujio; Sasa, Toshinobu; Obayashi, Hironari; Takei, Hayanori

Nuclear Engineering and Design, 403, p.112147_1 - 112147_17, 2023/03

https://doi.org/10.1016/j.nucengdes.2022.112147

It is important to evaluate the inventories and the release and transport behavior of the spallation products (SPs) in the Lead-Bismuth Eutectic (LBE) coolant system of Accelerator Driven System (ADS) for the safety studies of the radiological hazard both in the cases of normal operation and accident. University of Fukui and JAEA have been developing the computer analysis code TRAIL (Transport of RAdionuclides In Liquid metal systems) which predicts the time dependent behavior of SPs within the LBE coolant system of ADS for the wide range of operational events. The source term of both radioactive and stable SPs in the LBE coolant is given as input and the radioactive decay chain model for the radioactive SPs is implemented in the code to evaluate the effect of precursors on the SPs mobility. This paper presents the recent advancement status of the code development and the validation results comparing with the distribution data of volatile SPs in MEGAPIE spallation target.

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

Cappia, F.*; Tanaka, Kosuke; Kato, Masato; McClellan, K.*; Harp, J.*

Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05

https://doi.org/10.1016/j.jnucmat.2020.152076

Analysis of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya

Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02

https://doi.org/10.1080/00223131.2017.1384705

The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for U, Np, Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of Np. The C/E-1 values do not depend on the types of fuel rods (UO or UO-GdO) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.

Development of the prediction technology of cable disconnection of in-core neutron detector for the future high-temperature gas-cooled reactors

Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.041008_1 - 041008_5, 2016/10

https://doi.org/10.1115/1.4032597

Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.

Development of the prediction technology of cable disconnection of in-core neutron detector for the future high-temperature gas cooled reactors

Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.

Corrections to the Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

Suyama, Kenya; Mochizuki, Hiroki*

Annals of Nuclear Energy, 33(4), p.335 - 342, 2006/03

https://doi.org/10.1016/j.anucene.2005.11.009

The value of the burnup is one of the most important parameters of samples taken by post irradiation examination (PIE). In this study, concerning the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken at the Japan Atomic Energy Research Institute, the burnup values of the PIE samples were re-evaluated and the PIE data are re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. This analysis concludes that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected of 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is approximately 1% for PIE samples having the burnup of larger than 30 GWd/t. Comparison between calculation results using a single pin cell model and an assembly model is carried out. Because the both results agreed within a few percents, we concluded that the single pin cell model is suitable for the analysis of PIE samples and the underestimation of plutonium isotopes does not result from the geometry model.

Nondestructive observation of nuclear fuels and materials by using neutron radiography

Yasuda, Ryo; Matsubayashi, Masahito; Nakata, Masahito; Matsue, Hideaki; Nakanishi, Tomoko

Dai-5-Kai Hoshasen Ni Yoru Hihakai Hyoka Shimpojiumu Koen Rombunshu, p.31 - 34, 2005/02

Neutron radiography is an effectively nondestructive tool for inspection of irradiated nuclear fuels and materials. Neutron CT an neutron imaging plate methods, which are advanced techniques in the neutron radiography, enabled to obtain cross-section images and to evaluate an amount of the element compositions. This paper describes results of those methods using unirradiated fuels and materials and discussed the practicability of those methods to irradiated fuels and materials.

Development of test techniques for in-pile SCC initiation and growth tests and the current status of in-pile testing at JMTR

Ugachi, Hirokazu; Kaji, Yoshiyuki; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.319 - 325, 2005/00

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this conference, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack initiation, propagation and water chemistry, and the current status of in-pile SCC tests using thermally sensitized materials at JMTR.

Analysis of benchmark results for reactor physics of LWR next generation fuels

Kitada, Takanori*; Okumura, Keisuke; Unesaki, Hironobu*; Saji, Etsuro*

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 8 Pages, 2004/04

Burnup calculation benchmark has been carried out for the LWR next generation fuels aiming at high burnup up to 70 GWd/t with UO and MOX. Based on the submitted results by many benchmark participants, the present status of calculation accuracy has been confirmed for reactor physics parameters of the LWR next generation fuels, and the factors causing the calculation differences were analyzed in detail. Moreover, the future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.