Development and validation of sodium fire analysis code, ASSCOPS

; ; Tanabe, Hiromi; Ohno, Shuji; Miyake, Osamu;

PNC TN9410 97-030, 93 Pages, 1997/04

PNC-TN9410-97-030.pdf:2.2MB

A sodium fire analysis code, ASSCOPS(Analysis of Simultaneous Sodium Combustions in Pool and Spray) was developed coupling the computer codes of SPRAY-IIIM and SOFIRE-MIl to assess temperature-pressure transients resulting from sodium spray and pool combustions, simultaneously. The validation of ASSCOPS was conducted using the experimental results obtained from sodium spray fire experiments using 21 m vessel and the accuracy of calculated results was discussed. The following results were obtained: (1)Study under inert gas atmosphere. The comparison between analysis and experiment with regard to the pressure and the temperature showed a good agreement. (2)Study under air atmosphere. The comparison between analysis and experiment with regard to the pressure and the temperature also showed a good agreement. (3)Effects of parameter used in evaluating the design of Monju. The peak pressure and temperature obtained by the analysis overestimates the experimental results. From these results, it was concluded that the development and validation of ASSCOPS indicate a improvement on the burning and the heat transfer models in SPRAY-IIIM.

Progress report of the design study on a large reactor

; Hayashi, Hideyuki; ; ;

PNC TN9410 94-222, 355 Pages, 1994/07

PNC-TN9410-94-222.pdf:14.85MB

A design study on a large scale fast reactor was performed with focusing on enhancement of passive safety and capital cost reduction. The passive safety feature in the plant design of next generation fast reactors is one of the important subjects to be sought. In FY 1993, studies on 1300MWe class lage fast reactor were performed aiming at passive shutdown in a typical ATWS such as the unprotected loss of flow accident (ULOF). This report describes the core design, systems design, equipments design and the technical assessment in terms of the passive safety feature. It also includes the evaluation of the seismic as well as thermal capacity of reactor building of the large fast reactor.

Progress report of the design study on a large scale reactor

; Hayashi, Hideyuki; ; Yamaoka, Mitsuaki; ; ;

PNC TN9410 93-162, 494 Pages, 1993/07

PNC-TN9410-93-162.pdf:24.65MB

A design study on a large scale fast reactor was performed based on the results of the study on 600 MWe class "reactor vessel head access concept" plant in FY19901991. The objective of this study is to confirm quantitatively the applicability of "reactor vessel head access concept" to larger scale fast reactor plant as well as to establish the basic design such as core design, reactor structure, heat transport system, of 1300 MWe class "reactor vessel head access concept" plant and to evaluate its feasibility. In the former half period of FY1992, studies were performed aiming at evaluation of precision of core design, integrity of the structure and safety reflecting the final equipment design, feasibility of the concept of spent fuel storage, etc. While in the latter half period, added were the subjects on key technologies and plant concepts which enhance economy, safety, and social acceptance in order to ascertain the direction of a design study in the future and to contribute the realization of a commecial fast reactor.

None

; Hayashi, Hideyuki; ; ; ; ;

Donen Giho, (82), p.19 - 37, 1992/06

https://jopss.jaea.go.jp/pdfdata/PNC-TN1340-92-002.pdf

None

Safety aspect of future LMFBR design in Japan; PNC view

*

PNC TN9410 89-127, 33 Pages, 1989/08

PNC-TN9410-89-127.pdf:0.94MB

Safety aspects of future LMFBR design in Japan, comparing with those of experimental plant "JOYO" and prototype plant "MONJU", were shown. Safety aspects were discussed from the viewpoint of Controllability, Coolability and Containability. As a future plant of Japanese LMFBR, a plant design without secondary heat transport system using double walled tube steam generator was discussed. This paper was prepared for Japanese contribution to the Session VI of the IAEA international Workshop on the Safety of Nuclear Installations of the Next Generation and Beyond held at Chicago.

A Study of sodium iodide solubility in sodium; Testin sodium temperature range in 500C-800C

*; *; *; *; *

PNC TN9410 89-034, 56 Pages, 1989/02

PNC-TN9410-89-034.pdf:1.08MB

The saturated solubility of NaI in sodium was measured to be used for the safety analysis of the Fast Breeder Reactors. A saturated preparation vessel, and test apparatus was designed and preparation procedures for the study was established. The test was conducted in the sodium temperature range of 500C-800C. From the resu1ts obtained the two empirical equations saturated solubility have been derived: [logS = 8.557-3524/T(800C-661.4C)] [logS = 9.751-4640/T(661.4C-500C)] [S, S : saturated solubility (ppm)] [T : absolute temperature (K)] The above two equations of the saturated solubility represented test result with 1% errors. An intersection of the equations completely agreed with the melting point 661.4C of NaI. The results with the present ones by M, A, Bredig et al. agrees.

Verification and Validation of LMFBR Static Core Mechanics Codes Final Report an IWGFR Co-ordinated Research Programme

*; *

Final Report of the CRP on Intercomparison of LMFBR Core MEchanics Codes, 0 Pages, 1989/00

no abstracts in English

Preliminary results on experiments, analyses and evaluations performed by Reactor Technology Section, Experimental Reactor Division; Quarterly report Vol.5, No.1

*; *

PNC TN9410 88-096, 191 Pages, 1988/08

PNC-TN9410-88-096.pdf:19.56MB

This report summarizes preliminary results on experiments, analyses and evaluations performed by Reactor Technology Section, Experimental Reactor Division during April through June, 1988. An results described in the report were released with internal memoranda of Reactor Technology Section, before detailed analyses, evaluations, and/or discussions. Each result is classified into the following categories according to its content. (1)Results of measurements, analyses and evaluations for nuclear characteristics in JOYO. (2)Results of measurements, analyses and evaluations for in-core thermal-hydraulic characteristics. (3)Results of analyses and evaluations for core mechanics. (4)Results of measurements and analyses for the plant characteristics of JOYO. (5)Results of measurements and analyses of neutron flux, gamma ray and decay heat of JOYO. (6)Preparation of Construction Permit Amendment for the operational reliability tests of FBR fuel assembly. (7)Preparation of Construction Permit Amendment for spent fuel strage facility. (8)Results of study and development for analytical technic. (9)Production and/or arrangements of analyses codes and their manual. (10)Miscellaneous results. Tables classified by kind of work are added. After detailed discussions, analyses and evaluations, the final report for each program will be published, respectively.

Preliminary results on experiments, analyses and evaluations performed by reactor technology section, experimental reactor division; Quarterly report vol.4, No.3

*; *

PNC TN9410 88-011, 184 Pages, 1988/02

PNC-TN9410-88-011.pdf:20.19MB

This report summarizes preliminary results on experiments, analyses and evaluations pertormed by Reactor Technology Section, Experimental Reactor Division during october through December, 1987. All results described in the report were released with internal memoranda of Reactor Technology Section, before detailed analyses, evaluations, and/or discussions. Each result is classified into the following categories according to its content. (1)Results of measurements, analyses and evaluations for nuclear characteristics in JOYO. (2)Results of measurements, analyses and evaluations for in-core thermal-hydraulic characteristics. (3)Results of analyses and evaluations for official permission. (4)Results of analyses and evaluations for core mechanics. (5)Results of measurements and analyses for the plant characteristics of JOYO. (6)Results of measurements and analyses of neutron flux, gamma ray and decay heat of JOYO. (7)Production and/or arrangements of analyses codes and their manual. (8)Preparation of Construction Permit Amendment for RTCB (Run To Claddlng Breach) test. (9)Miscellaneous results. Tables classified by kind of work are added. After detailed discussions, analyses and evaluations, the final report for each program will be published, respectively.

Preliminary results on experiments, analyses and evaluations performed by reactor technology section, experimental reactor division; Quarterly report Vol.4, No.1

*; *

PNC TN9410 87-103, 160 Pages, 1987/09

PNC-TN9410-87-103.pdf:9.26MB

This report summarizes preliminary results on experiments, analyses and evaluations performed by Reactor Technology Section, Experimental Reactor Division during April through June, 1987. All results discribed in the report were released with internal memoranda of Reactor Technology Section, before detail analyses, evaluations, and/or discussions, Each result is classified into the following categories according to its content. (1)Results of measurements, analyses and evaluations for core characteristics of JOYO. (2)Results of analyses and evaluations for transfer plan to J2 core. (3)Results of analyses and evaluations for official permission. (4)Results of analyses and evaluations for core mechanics. (5)Results of measurements and analyses for natural circulation test in JOYO MK-II core. (6)Results of measurements and analyses for the plant characteristics of Expermiental Fast Reactor "JOYO". (7)Results of measurements and analyses of neutron flux and gamma ray of JOYO. (8)Production and/or arrangements of analysis codes and their manuals. (9)Miscellaneous results. Tables classified by kind of work are added. After detail discussions, analyses and evaluations, the final report for each program will be published, respectively.