Experimental investigations of the hydraulic field in wire-wrapped LMFBR core assemblies

Roidt, R.M.; Carelli, Mario D.; Markley, R.A.

doi:10.1016/0029-5493(80)90035-7

Cited by 40 publications

(5 citation statements)

References 4 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A forced flow inter-channel mixing model applicable to LMFBR fuel assemblies containing helical wire spacer systems that can be incorporated into COBRA-II and COBRA-III B lumped-parameter computer codes was presented by Ginsberg [1]. An investigation of the hydraulic behavior of wire-wrapped fuel and blanket assemblies in an air flow test facility was conducted by Roidt et al [2]. Experimental results were discussed in detail below.…”

Section: Introductionmentioning

confidence: 99%

Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium

Hata

Liu

Nakajima

2019

Nuclear Engineering and Design

View full text Add to dashboard Cite

Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium was numerically analyzed on the following four effects, the influences of surface heat flux, gap between helical wire and heated cylinder surface, pitch of helical wire and helical wire diameter on heat transfer. The unsteady laminar three dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the numerical solution reaches a steady-state. The PHOENICS code was used for the calculation considering the temperature dependence of thermo-physical properties concerned. The heated cylinder for diameter (D=7.6 mm), heated length (L=200 mm) and L/d (=26.32), and the helical wire spacers for gaps (0 to 1 mm), pitches (100 to 200 mm), helical wire diameters (d=1 to 2 mm) and height (H=470 mm) were used in this work. The surface heat fluxes for each cylinder with helical wire spacer were equally given for a modified Rayleigh number, (Ra f,L ) w-s , ranging from 3.05910 4 to 3.14710 7 (q=110 4 to 710 6 W/m 2 ) in liquid temperature (T L =673.15 K). The values of gap were given from 0 to 1 mm at d=1.2 mm and a pitch of 165 mm, those of pitch were given from 100 to 200 mm at d=1.2 mm and a gap of 0.2 mm and those of wire diameter were given from 1 to 2 mm at a gap of 0.2 mm and a pitch of 165 mm on each cylinder with helical wire spacer. The spatial distributions of local and average Nusselt numbers, (Nu ,z ) w-s and (Nu av ) w-s , on each cylinder with helical wire spacer were clarified. The values of average Nusselt number, (Nu av ) w-s , for a vertical cylinder with helical wire spacer were calculated to examine the effects of (Ra f,L ) w-s , gap, pitch and helical wire diameter on heat transfer. The correlation for (Nu av ) w-s for a vertical cylinder with helical wire spacer above mentioned including the effects of (Ra f,L ) w-s , gap, pitch and helical wire diameter was developed.The correlation can describe the theoretical values of (Nu av ) w-s for a vertical cylinder with helical wire spacer within -2.74 to 1.63 % differences.

show abstract

Section: Introductionmentioning

confidence: 99%

Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium

Hata

Liu

Nakajima

2019

Nuclear Engineering and Design

View full text Add to dashboard Cite

show abstract

“…The data of Okamoto (1970), Hines (1971), Collingham (1971), Skok (1973) were taken from Khan's (1974) reduction results for e* by the old DRV * code, which were transferred to e, by equation (5 Table 5.1 is 20 -30%. H/D <a Roidt (1980) 52.0 9 Roidt(1980) 7.9 A Fontana(1974) 52.0 V Okamoto(1970) 40.5 • Hines(1971 24.0 • Collingham 52.0 (1971) a Skok(1973) 7-pin 0.06 0.08 0.1 0.15 0.2 0.3 0.4 c/D By matching the transition data of our 37-pin results and Chiu's two 61-pin results, a value of 2/3 for X was found to be optimum. The empirical constants C , and C " are then ^ mltr m2tr C ., = C ., + (C ., -C .,)\p.^ (5.40) mltr raiL miT raiL i…”

Section: Bundle Friction Factor Constantsmentioning

confidence: 99%

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

Todreas¹,

Cheng²,

Basehore³

1984

View full text Add to dashboard Cite

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenoraenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

show abstract

“…Interaction between channels occurs due to lateral pressure difference, spacer, [ 7 ] and turbulence. Interaction due to turbulence is treated differently by sub‐channel codes.…”

Section: Introductionmentioning

confidence: 99%

Estimation of turbulent energy mixing factor in PWR sub‐channel by DNS

Singh,

Mukhopadhyay,

Khakhar

et al. 2024

Can J Chem Eng

View full text Add to dashboard Cite

Turbulent mixing within sub‐channels plays a crucial role in understanding the thermal hydraulics of reactor channels. It serves as an empirical parameter in sub‐channel analysis and has long been a challenge in the nuclear industry. Conducting experiments in this context is challenging due to the stringent requirement of maintaining pressure balance among sub‐channels to prevent convection effects. Fortunately, direct numerical simulation (DNS) is emerging as an invaluable tool for addressing this persistent issue. DNS enables the direct computation of turbulent mixing by analyzing fluctuating lateral velocities, offering a more profound understanding of the underlying phenomena. In this study, DNS was conducted at six Reynolds numbers ranging from 17,640 to 1.5 × 105 in pressurized water reactor (PWR) geometry to investigate the lateral mixing driven by turbulence. By studying intricate mechanisms governing the turbulent mixing, the valuable insights into reactor thermal performance and safety are provided. Furthermore, a correlation for turbulent mixing of energy based on the DNS data has been derived, enhancing our ability to model and predict this critical aspect of reactor behaviour. Additionally, this paper explores temperature fluctuations occurring at the fuel rod surface due to turbulence. A probabilistic distribution for temperature fluctuation under specific reactor conditions is presented.

show abstract

Experimental investigations of the hydraulic field in wire-wrapped LMFBR core assemblies

Cited by 40 publications

References 4 publications

Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium

Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

Estimation of turbulent energy mixing factor in PWR sub‐channel by DNS

Contact Info

Product

Resources

About