Loss of Forced Coolant (LOFC) Project
Ongoing
Joint project

JAEA  High-temperature Engineering Test Reactor. Photo: Japan Atomic Energy Agency (JAEA)

 

The Committee on the Safety of Nuclear Installations (CSNI) Task Group on Advanced Reactor Experimental Facilities (TAREF) was established in 2007 and issued a report on the experimental facilities for gas-cooled reactor safety studies in June 2009. The group concluded that actions should be taken to develop an international project centred on the high-temperature engineering test reactor (HTTR) capabilities and focused on the safety issues identified by the TAREF. The Japanese HTTR operated by the Japan Atomic Energy Agency (JAEA) was identified as a unique resource and the  only experimental high-temperature gas-cooled reactor (HTGR) facility available at the time in the OECD Nuclear Energy Agency member countries. It is a graphite-moderated, helium-cooled reactor that can reach temperatures as high as 1600°C in some transient conditions. The loss of forced coolant (LOFC) experiments planned to study effects of reactor cavity cooling system (RCCS) performance reduction are highly relevant for safety assessments of advanced reactors such as high temperature reactor (HTR).

The present programme is formulated to investigate safety issues and specifically the anticipated transient without scram (ATWS) with occurrence of reactor re-criticality. The programme is devised to maximise the information deliverables for code validation for some of the most important safety aspects about reactor kinetics, core physics and thermal hydraulics. The experimental programme consists of three test cases whose results comparison will provide the incremental performance availability within the vessel cooling system (VCS) range. The experimental programme provides an experimental database, which will be used to validate code predictive capability and accuracy of models. Phenomena coupled between reactor core physics and thermal hydraulics are to be investigated. The experimental programme and associated analytical activities should foster the creation of a group among NEA member countries which share the need to maintain or improve the technical competence in reactor physics and thermal-hydraulics for safety evaluations of advanced gas cooled nuclear reactor.

The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MWt. The reactor outlet coolant temperature is 850˚C at the rated operation and 950˚C in high temperature test operation, respectively. The LOFC tests are performed at the rated operation mode. The HTTR consists of a reactor core, cooling system, engineered safety systems, instrumentation and control system, etc.

The nuclear instrumentation system of the HTTR is composed of the wide range monitoring system (WRMS) and the power range monitoring system (PRMS). The reactor control system is designed to assure high stability and reasonably damped characteristics against the various disturbances during the operation. The main control system of the HTTR consists of the operational mode selector, the reactor power control and plant control systems. The safety protection system consists of the reactor protection and engineered safety features actuating systems.

LOFC tests are initiated by tripping all three Helium Gas Circulators (HGCs) of the HTTR while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The test falls into ATWS with occurrence of reactor re-criticality. The test are conducted with and without active function of the VCS.

The Japan Atomic Energy Agency (JAEA) has conducted three tests in the HTTR. The first test was completed in December 2010. The project was put on hold after the 11 March 2011 Great Eastern Japan Earthquake and tsunami while active discussions were conducted between the JAEA and the regulator on the requirements for the safe operation of the HTTR. In June 2020, after implementing modifications requested by the regulator, the JAEA was given permission to restart the reactor, which it did in July 2021. A second test was performed in January 2022 and the last test was completed in March 2024.

Project members have decided to perform a code benchmark exercise on one of the tests. This multiphysics code-to-data benchmark provides an opportunity to assess modelling capabilities for prismatic HTGRs. The results of this benchmark will illustrate the degree to which each participant’s tools capture the neutronics and thermal hydraulics effects that dictate the progression of the experiment and can demonstrate the appropriateness of the codes for similar modelling. The project was extended until March 2027 to complete all experimental and modelling activities and related reports.

Related activities: Generation IV International Forum Publications

 

LOFC members' area (legacy workspace) (password protected | reminder)

LOFC members' area (new workspace) (password protected | reminder)

Participants

Czechia, France, Germany, Hungary, Japan, Korea and the United States

Project period

April 2011 to March 2027

Budget

EUR 3 million