Report a Safety Concern
Search

The U.S. Nuclear Regulatory Commission is in the process of rescinding or revising guidance and policies posted on this webpage in accordance with Executive Order 14151 Ending Radical and Wasteful Government DEI Programs and Preferencing, and Executive Order 14168 Defending Women From Gender Ideology Extremism and Restoring Biological Truth to the Federal Government. In the interim, any previously issued diversity, equity, inclusion, or gender-related guidance on this webpage should be considered rescinded that is inconsistent with these Executive Orders.

Home NRC Library Document Collections NUREG-Series Publications Publications Resulting from International Agreements

Assessment of TRACE 5.0 against ROSA Test 6-1, Vessel Upper Head SBLOCA (NUREG/IA-0245)

On this page:

Download complete document

Publication Information

Manuscript Completed: June 2010
Date Published: April 2011

Prepared by:
S. Gallardo, V. Abella, G. Verdú

Universidad Politécnica de Valencia
ETSII
Camí de Vera s/n
46021 Valencia, SPAIN

Antony Calvo, NRC Project Manager

Prepared For:
Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the International Code Assessment and Maintenance Program (CAMP)

Published by
U.S. Nuclear Regulatory Commission

Availability Notice

Abstract

The purpose of this work is to provide an overview of the results obtained in the simulation of a pressure vessel upper head Small Break Loss-Of-Coolant Accident (SBLOCA) under the assumption of total failure of High Pressure Injection System (HPIS) in the Large Scale Test Facility (LSTF) via the thermal-hydraulic code TRACE5.

The work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Test 6-1 simulated a PWR pressure vessel upper-head SBLOCA with a break size equivalent to 1.9% of the cold leg break. The break size and core uncover caused the primary depressurization. When the accident management (AM) action was initiated by fully opening the steam generator relief valves (detection of high core exit temperature of 623 K) primary pressure was lower than the steam generator secondary-side pressure. A detailed model has been developed following these assumptions.

Results of the simulation with TRACE5 are compared with the experimental in several graphs, observing an acceptable general behaviour in the entire transient. In conclusion, this work represents a good contribution for assessment of the predictability of computer codes such as TRACE5.

Page Last Reviewed/Updated Tuesday, March 09, 2021