Conference Proceedings Paper
Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmarks Database
NURETH-14 - 2011 September 25-30
Maria Avramova (The Pennsylvania State University)
Diana Cuervo (Technical University of Madrid)
The need to refine models for best-estimate calculations, based on good-quality experimental
data, has been expressed in many recent meetings in the field of nuclear applications. The
modeling needs arising in this respect should not be limited to the currently available
macroscopic methods but should be extended to next-generation analysis techniques that focus
on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics
modeling was developed by the Nuclear Power Engineering Corporation (NUPEC),
Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and
departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups.
Considering the reliability not only of the measured data, but also other relevant parameters such
as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied
the first substantial database for the development of truly mechanistic and consistent models for
boiling transition and critical heat flux.
Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the
U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and
summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The
international benchmark activities have been conducted in cooperation with the Nuclear Energy
Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan
Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available
the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based
on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the
data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for
another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle
Tests (PSBT) benchmark.
This paper presents an application of the joint Penn State University/
Technical University of
version of the well-known subchannel code COBRA-TF, namely CTF, to the
critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT
and PSBT benchmarks.
Individual Conference-Paper Copies (Electronic Where Available):
- For CNS members, the first 5 copies per calendar year are free, and additional copies are $10 each.
- For non-members, the price is $25 for the first Conference-paper copy in a request, and $10 each for additional copies of papers in the same conference and in the same request.
- Contact the CNS office to order reprints.