• Home
• About Us
  • What's New
  • Press Room
  • Contacts
  • Vacancies
  • Practical Info
  • Register
  • Web Links
  • List of Acronyms
• Work Areas
  • Nuclear Safety
  • Radioactive Waste
  • Radiological Protection
  • Nuclear Development
  • Nuclear Science
  • Nuclear Law 
  • Joint Projects
  • Sustainable Development
  • Civil Society
• Data Bank
  • Nuclear Data 
  • TDB Project 
  • Computer Programs
• Publications
  • By Subject 
  • Recently Published
  • NEA News
  • Policy Papers
• Press Room
• Search

[Program Service | Recent programs]

CSNI0044 ROSA-III/971. (Abstract last modified 29-JUN-1992)


ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient

---

Abstract Table of Contents:

NAME | COMPUTER | PROBLEM | SOLUTION | RESTRICTIONS | CPU | FEATURES | AUXILIARIES | STATUS | REFERENCES | REQUIREMENTS | LANGUAGE | OPERATING SYSTEM | OTHER RESTRICTIONS | AUTHOR | MATERIAL | CATEGORIES |

---

1. NAME OR DESIGNATION - ROSA-III/971.


2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

To request or retrieve programs click on the one of the active versions below. A password and special authorization is required.

Explanation of the status codes.

Program-name         Package-ID  Status
ROSA-III/971         CSNI0044/01 Report

Machines used:
Package-ID     Orig.Computer                       Test Computer
CSNI0044/01    Many Computers                      

3. DESCRIPTION OF TEST FACILITY - ROSA-III is a 1/124 scaled down test facility with electrically heated core designed to study the response of engineered safety features to loss-of-coolant accidents in in commercial BWR. It consists of the following, fully instrumented subsystems:
(a) the pressure vessel with a core simulating four half-length fuel assemblies and control rod;
(b) steam line and feed water line, which are independent open loops;
(c) coolant recirculation system, which consists of two loops provided with a recirculation pump and two jet pumps in each loop;
(d) emergency cooling system, including HPCS, LPCS, LPCI, and ADS.


4. DESCRIPTION OF TEST - Run 971 simulated a BWR LOSS of off-site power transient. The core scram was assumed to occur at 6 seconds after the transient initiated by the turbine trip. HPCS failure was assumed. After ADS started, the upper half of the core was uncovered by steam. The core was reflooded by LPCS alone.


5. EXPERIMENTAL LIMITATIONS OR SHORTCOMINGS -


6. PHENOMENA TESTED - Core thermal hydraulics; parallel channel effects and instabilities; void collapse and temperature distribution during pressurization; critical power ratio.


7. SPECIAL FEATURES OF EXPERIMENT -


8. COUNTERPART EXPERIMENTS -


9. STATUS
CSNI0044/01: 29-JUN-1992 Report Only


10. REFERENCES  - 
CSNI0044/01:
- Y. ANODA, K. TASAKA, H. KUMAMARU, M. SHIBA:
ROSA-III System Description for Fuel Assembly No. 4
JAERI-M 9363, February 1981
- H. Nakamura et al.:
Loss of Off-Site Power Test of Rosa-III (Run 971 with HPCS Failure)
JAERI-M 89-130, September 1989


11. TEST DESIGNATION - 971


12. PROGRAMMING LANGUAGE(S) USED -
CSNI0044/01: 


15. ESTABLISHMENT -
     Japan Atomic Energy Research Institute
     TOKAI-MURA
     Naka-gun, Ibaraki-ken 319-11
     JAPAN


16. MATERIAL AVAILABLE -
CSNI0044/01:
Report ROSA-III Test 971                   


17. CATEGORIES  - 

- Y. Integral Experiments Data, Databases, Benchmarks


Keywords: DATA, LOSS-OF-COOLANT ACCIDENT, TRANSIENTS

---

Home - About Us - Work Areas - Data Bank - Publications - Press Room - List of acronyms - Search