A Review of the Scaling Study of the CANDU-6 Moderator Circulation Test Facility

Bo Wook Rhee^*, Hyoung Tae Kim

Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute (KAERI), Daejeon, Republic of Korea.

DOI: 10.4236/jpee.2014.29010                

Abstract

Following the previous relevant works [1]-[3], a scaling analysis is performed to derive a set of scaling criteria which were thought to be suitable for reproducing the major thermal-hydraulic phenomena in a scaled-down CANDU moderator tank similar to that in a prototype power plant during a full power steady state condition. The objective of building a scaled-down moderator tank is to generate the experimental data necessary to validate the computer codes which are used to analyze the accident analysis of CANDU-6 plants. The major variables of interests in this paper are moderator flow velocity and temperature of the moderator which is D2O inside the moderator tank during a steady state and transient conditions. The reason is that the local subcooling of the moderator is found to be a critical parameter determining whether the stable film boiling can sustain on the outer surface of the calandria tube if the contact of overheated pressure tube and cold calandria tube should occur due to pressure tube ballooning during LBLOCA with ECC injection failure [4]. The key phenomena involved include the inlet jet development and impingement, buoyancy force driven by the moderator temperature gradient caused by non-uniform direct heating of the moderator, and the pressure drop due to viscous friction of the flow across the calandria tube array. In this paper, the previous researches are reviewed, some concerns or potential problems associated with them implied by comparing CFD analyses results between the CANDU-6 moderator tank and 1/4 scaled-down test facility are described, and as a way to examine the assumption of the scaling analysis is true an order-of-magnitude analyses are performed. Based on the results of these analyses the assumption of neglecting  and  terms cannot be justified for the power of 0.5 MW and 1.566 MW for the 1/4 scaled-down facility. Further investigation is thought to be necessary to confirm this result, i.e. if the scaling of the previous work1 is justifiable by some other independent analyses.

Keywords

CANDU-6, Thermal-Hydraulic Phenomena, Moderator Tank, Experimental Test Facility, Scaling Analysis

Conflicts of Interest

The authors declare no conflicts of interest.

References

[1]	Khartabil, H.F., et al. (2000) Three-Dimensional Moderator Circulation Experimental Program for Validation of CFD Code MODTURC_CLAS. 21th CNS Nuclear Simulation Symposium.
[2]	Rhee, B.W. and Kim, H.T. (2012) Derivation of the Scaling Criteria for a Small Scale CANDU Moderator Test Facility. Transactions of the KNS Spring Meeting, Jeju, 17-18 May 2012.
[3]	Rhee, B.W., Kim, H.T., Park, S.K., Cha, J.E., Choi, H.-L. (2013) A Scaling Analysis of a CANDU-6 Moderator Tank Scaled-Down Test Facility. Proceedings of the 21st International Conference on Nuclear Engineering, ICONE21, 29 July-2 August2013, Chengdu.
[4]	Gillespie, G.E., Moyer, R.G. and Thompson, P.D. (1982) Moderator Boiling on the External Surface of a Calandria Tube in a CANDU Reactor during a Loss-of-Coolant Accident. Proceedings of the International Meeting on Thermal Nuclear Reactor Safety, 29 August-2 September 1982, Chicago, USNRC, Report No. NUREG/CP-00, AECL-7604.