Published 2011 | Version v1
Miscellaneous Metadata-only

The integral test facility Karlstein - INKA

Creators

  • 1. AREVA NP Offenbach (Germany)

Description

The INKA (INtegral Test Facility KArlstein) test facility was designed and erected to test and demonstrate performance of the passive safety systems of KERENA™, the new AREVA Boiling Water Reactor (BWR) design. The experimental program within the KERENA™ development program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests will be performed to simulate transients and LOCA (Loss of Coolant Accident) scenarios at the INKA test facility. These tests will test and demonstrate the interaction between the passive components/systems and demonstrate their ability to perform their design function. For the integral tests, the Passive Pressure Pulse Transmitter will be included. The INKA test facility represents the KERENA™ Containment with a volume scaling of 1:24. Component heights and levels are full scale in order to match the driving forces for natural circulation. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein - a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The drywell of the INKA containment is divided into two compartments and connected to the wetwell (Pressure Suppression System) via a full scale vent pipe. Therefore, the INKA pressure suppression system meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios starting with the initiating event - for example pipe rupture. At INKA a full train of passive safety systems is available. INKA is also able to simulate the functions of active safety system such as containment heat removal. Therefore accident scenarios relevant to modern Gen III as well as for operating Gen II design can be simulated in order to validate system and containment codes. (author)

Part of:
The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14)

Additional details

Publishing Information

Publisher
Canadian Nuclear Society
Imprint Place
Toronto, Ontario (Canada)
ISBN
978-1-926773-05-6
Imprint Title
The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14)
Imprint Pagination
766 Megabytes
Journal Page Range
[13 p.]

Conference

Title
14. International Topical Meeting on Nuclear Reactor Thermalhydraulics
Acronym
NURETH-14
Dates
25-30 Sep 2011
Place
Toronto, Ontario (Canada)

INIS

Country of Publication
Canada
Country of Input or Organization
Canada
INIS RN
47071301
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference, Non-conventional Literature
Descriptors DEI
BWR TYPE REACTORS; CORE FLOODING SYSTEMS; ECCS; ENGINEERED SAFETY SYSTEMS; LOSS OF COOLANT; REACTOR ACCIDENT SIMULATION; TEST FACILITIES
Descriptors DEC
ACCIDENTS; ECCS; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Notes
Paper NURETH14-357, 4 refs., 8 figs.