Phenomenology of BWR fuel assembly degradation

Kurata, Masaki; Barrachin, M.*; Haste, T.*; Steinbrueck, M.*

Journal of Nuclear Materials, 500, p.119 - 140, 2018/03

https://doi.org/10.1016/j.jnucmat.2017.12.004

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (BC) control rods potentially generate far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with BC control rods or control blades have shown early failure, liquefaction, relocation and oxidation of BC starting at temperatures around 1250C, well below the significant interaction temperatures of UO-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

A Recent experimental program to evidence in-vessel retention by controlled material relocation during core disruptive accidents of sodium-cooled fast reactors

Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Zuev, V. A.*; Ganovichev, D. A.*; Kolodeshnikov, A. A.*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

Molten fuel discharge through control rod guide tubes (CRGTs) is a key process that dominates the termination of core disruptive accidents of sodium-cooled fast reactors, since fuel dispersion from the core contributes to the achievement of both deeper subcriticality in the degraded core and formation of coolable debris bed. Within a framework of a collaborative research program between Japan Atomic Energy Agency and National Nuclear Center of the Republic of Kazakhstan, called EAGLE program, a new experimental program has been started with out-of-pile experiments to clarify the fuel discharge through CRGTs. This paper presents the status of the new program, including experimental results obtained so far.

Preventive measures for the recurrence of control rod insertion failure in Japan Research Reactor No.4 (JRR-4)

JRR-4 Operation Division; Research Reactor Utilization Division

JAERI-Tech 2005-042, 58 Pages, 2005/07

JAERI-Tech-2005-042.pdf:5.4MB

Japan Research Reactor No.4 (JRR-4) was shut down manually, due to the control rod insertion failure occurred during the rated power (3,500kW) operation on June 10, 2005. It became evident by the investigation that a screw bolt at the control rod support got loose and blocked the control rod insertion. The failure was recovered through replacement with the new screw bolt. Considering the importance of this event, we decided to inspect all screw bolts over the core that may cause a control rod insertion failure. Furthermore, we decided to carry out periodical inspection about these screw bolts whether they were tightened enough or not. This report describes the result of inspection carried out as the preventive measures.

Detail analysis for a control rod worth of the Gas Turbine High Temperature Reactor (GTHTR300)

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

JAERI-Tech 2002-087, 83 Pages, 2002/11

JAERI-Tech-2002-087.pdf:3.47MB

no abstracts in English

Practical use of control rod calibration system with the inverse kinetices method

Yamanaka, Haruhiko; Hayashi, Kazuhiko; Motohashi, Jun; Kawashima, Kazuhito; Ichimura, Toshiyuki; Tamai, Kazuo; Takeuchi, Mitsuo

JAERI-Tech 2001-084, 110 Pages, 2002/01

JAERI-Tech-2001-084.pdf:10.15MB

no abstracts in English

Design study on Reduced-Moderation Water Reactor (RMWR)

Okubo, Tsutomu; Iwamura, Takamichi; Yamamoto, Kazuhiko*; Okada, Hiroyuki*

Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.571 - 574, 2002/00

Based on the experienced light water reactor technology, conceptual design studies on advanced water-cooled reactors have been performed. They are named “Reduced-Moderation Water Reactor" (RMWR) with the high conversion ratio more than 1.0 and the negative void reactivity coefficients. Several concepts have been successfully established for them based on the neutronics calculations. Based on these concepts, detailed investigations on such as plutonium multiple recycling and control rod planning have been performed as well as improvement on core performances. Through these detailed core design investigation, the feasibility of those designs has been confirmed step by step.

Development of in-vessel type control rod drive mechanism for marine reactor

Ishida, Toshihisa; Imayoshi, Sho*; Yoritsune, Tsutomu; Nunokaya, Hiroshi*; Ochiai, Masaaki; Ishizaka, Yuichi*

Journal of Nuclear Science and Technology, 38(7), p.557 - 570, 2001/07

https://doi.org/10.3327/jnst.38.557

no abstracts in English

The Development of the measurement technique of the control rod reactivity worth with the inverse kinetics method considering the influence of the steady neutron source

Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

JAERI-Tech 2000-054, 51 Pages, 2000/09

JAERI-Tech-2000-054.pdf:2.23MB

no abstracts in English

Analyse on the BFS critical experiments; An analysis on the BFS-62-1 assembly

Sugino, Kazuteru; Iwai, Takehiko*;

JNC TN9400 2000-098, 182 Pages, 2000/07

JNC-TN9400-2000-098.pdf:5.74MB

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0 fuel surrounded by the U0 blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....

Research and development on very small reactor for heat supply

Nakajima, Nobuya; Kusunoki, Tsuyoshi; Odano, Naoteru; Ochiai, Masaaki

Dai-7-Kai Doryoku Enerugi Gijutsu Shimpojiumu Koen Rombunshu (00-11), p.225 - 228, 2000/00

no abstracts in English