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


Presented at:
NURETH-14
2011 September 25-30
Location:
Toronto, Canada
Session Title:
O8-3 OECD Thermalhydraulics Benchmarks

Authors:
Maria Avramova (The Pennsylvania State University)
Diana Cuervo (Technical University of Madrid)
  

Abstract

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

Madrid (UPM)

 

 

 

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.

 

 

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