Test methods for robots for nuclear emergency response and decommissioning; Tests for moving performances of robots (JAEA-TM-0004 and JAEA-TM-0005)

Kawabata, Kuniaki; Yamada, Taichi; Abe, Hiroyuki*

JAEA-Technology 2021-021, 30 Pages, 2021/11

JAEA-Technology-2021-021.pdf:2.55MB

This report describes the test procedures for performance evaluation of remotely operated robot utilized for nuclear emergency responses and decommissioning that provide to compare among the robot's performances quantitatively and relatively. After the accident at Fukushima Daiichi Nuclear Power Station of the Tokyo Electric Power Company Holdings Inc. (FDNPS) occurred, remotely operated robots have been deployed and utilized in the response tasks. Such post-accident work experiences and lessons learned are very valuable for developing the robots in the future. Therefore, we were motivated to develop the test methods for performance evaluation of the robot by referring with such experiences and lessons. In recent decommissioning tasks, reconnaissance on the distribution and status of nuclear fuel debris inside the Primary Containment Vessel (PCV) have been carried out. The insertion and deployment of robots into PCV were carried out through a penetration pipe with small diameter to prevent the scattering of radioactive materials. According to the authors' survey on such works have carried out in Units 1 and 2 of FDNPS, in order to carry out the reconnaissance work by the robot deployed into the PCV, it was clarified that the robots are required to run freely on the floor located below the exit of the penetration pipe and run freely on the inclined surface located below the exit of the pipe. This document describes two test procedures for performance evaluation of the robot connected with the cable such as running on the floor after being deployed through a penetration pipe and running on the inclined surface after being deployed through a penetration pipe. Typical course layout and the demonstration of test running are also illustrated for the references.

Naraha Remote Technology Development Center and robots for nuclear disaster

Kawatsuma, Shinji

Dekomisshoningu Giho, (54), p.24 - 33, 2016/09

It has passed more than five years than Tokyo Electric Power Company's Fukushima Daiichi NPPs accidents occurred by huge tsunamis caused by the earthquake in Pacific Ocean Coast of North East District on March 11, 2011. It is very hard for workers to enter and stay for long time to work for decommissioning, because the radiation dose rate in the reactor buildings is too high to extremely high caused by radioactive materials released. Then the Naraha Remote Development Center has been constructed and taken into full operation in April 2016, which center would accelerate the development of remote technologies conducting decommissioning on the behalf of workers. The center is developing robot simulator system and robot performance testing method which could support developing remote operating equipment and devices. Also the center is preparing and operating remote equipment and devices for nuclear emergency response.

Development of an on-site plant analyzer (1); Development of a GUI for building plant models for analyzes and retrieval of real-time plant data

JNC TN4400 2000-002, 33 Pages, 2000/06

JNC-TN4400-2000-002.pdf:5.22MB

An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.

Development of a standard database for FBR core nuclear design (XI); Analysis of the experimental fast reactor "JOYO" MK-I start up test and oparation data

; Numata, Kazuyuki*

JNC TN9400 2000-036, 138 Pages, 2000/03

JNC-TN9400-2000-036.pdf:10.16MB

Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.

Hydrogen and tritium behaviour in Monju; Validation of an analysis code for tritium transport in fast reactor system, TTT, and estimation for Monju full power operation in future

;

JNC TN4400 99-002, 192 Pages, 1999/03

JNC-TN4400-99-002.pdf:7.27MB

The tritium transport analysis code, TTT, has been validated using data from the low power test of Monju, and then its behaviour at along term full power operation of Monju in future has been estimated, when the estimated transport and distribution of tritium in the reactor system has been also compared with the result in Joyo and Phenix, which had been already experienced long term operations. The TTT code had been develpped using the tiritium and hydrogen transport model proposed by R. Kumar, ANL, and had been applied to the evaluation in Monju design work. After then, futhermore, the code has been improved using the data from long term operation of Joyo with MK-II core, and in this work the code has been validated for the first time for Monju data. The results from this work are as follows; (1)Comparison of the best fitted tritium source rates from cores in Joyo, Phenix and Monju makes an estimation of the major source from control rods, (2)The calculated tritium concentration in each medium for cooling and its change is a reasonable agreement to the measured, C/E=1.1, (3)The cover gas transport model cosidering isotopic exchange of H and H can reproduce reasonably the measured concentration distirbution of tritium in sodium and cover gas, (4)The tritium concentration in secondary sodium of Monju was about l/50 times as much as the primary one, which shows the acceraration effect on cold tarapping of tritium due to coprecipitation with permeated hydrogen through Evaporater (EV) heat conduction tube walls. The tritium cold trapping efficiency was estimated to be 1 for coprecipitation with hydrogen and 0.3 for isotopic exchange, respectively, (5)Tritium transport and distribution for along term full power operation of Monju in future was estimated, which could involve a excess factor to 4 at the maximum. The tritium concentration in sodium and Steam Generator (SG) water will be substantially saturated after somthing like 10 years full power operation, ...

None

Hirata, Yoichi*; *

PNC TJ7439 98-003, 171 Pages, 1998/08

PNC-TJ7439-98-003.pdf:15.85MB

None

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*; Toida, Masaru*; Shiogama, Yukihiro*; *; Yasui, Shingo*; Fukazawa, E.*; Tanaka, M.*

PNC TJ1100 98-003, 204 Pages, 1998/02

PNC-TJ1100-98-003.pdf:9.91MB

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Application of a Phoswich detector for simultaneous counting of - and ()-rays in a rotating drum-cell type monitor

Usuda, Shigekazu; Yasuda, Kenichiro; Sakurai, Satoshi; Takahashi, Toshiyuki; Gunji, Hideho*; P.Howarth*

INMM 39th Annual Meeting Proceedings (CD-ROM), 27, 6 Pages, 1998/00

no abstracts in English

None

*; *; *

PNC TJ7439 96-005, 83 Pages, 1996/12

PNC-TJ7439-96-005.pdf:1.52MB

no abstracts in English

None

Kanno, Takeshi; Fujita, Tomoo; Sugita, Yutaka

PNC TN8410 95-179, 19 Pages, 1995/05

PNC-TN8410-95-179.pdf:0.67MB

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