Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

Kalimullah, M.; Olson, Arne; Feldman, E.E.; Hanan, N. A.; Pham, Son H.

doi:10.2172/1490817

Cited by 4 publications

(5 citation statements)

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“…This capability [20] is one of the options available in the code for estimating CHF, and is applicable only at near atmospheric pressure for coolant mass flux less than 400 kg/m 2 -s.…”

Section: Chfr For Low Flow and Atmospheric Pressurementioning

confidence: 99%

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

Kalimullah,

Olson,

Feldman

et al. 2023

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“…This capability [20] is one of the options available in the code for estimating CHF, and is applicable only at near atmospheric pressure for coolant mass flux less than 400 kg/m 2 -s.…”

Section: Chfr For Low Flow and Atmospheric Pressurementioning

confidence: 99%

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

Kalimullah,

Olson,

Feldman

et al. 2023

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“…The verification and validation (V&V) of the 16 capabilities of PLTEMP/ANL V4.3 (the current version) identified by research reactor analysts as frequently used in their thermal-hydraulic analysis is described in detail in revision 2 of the code V&V report [46], and a verification of the recently implemented heat transfer method with lateral conduction along fuel plate width, done by comparing the code-calculated temperature distribution with a closed-form analytical solution, is given in Appendix XV of this users guide.…”

Section: Verification and Validationmentioning

confidence: 99%

“…Equations (45) to (46) describe what goes to collect in a given sub-channel M due to mixing. Based on these equations, one can write equations for how the flow of a given sub-channel m splits into different sub-channels.…”

Section: Xiii2 Sub-channel Flow Mixing Modelmentioning

confidence: 99%

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A User’s Guide to the PLTEMP/ANL Code (V.4.3)

Kalimullah

Olson

Feldman

2021

Self Cite

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“…In addition, the PLTEMP/ANL code was also applied for modeling and simulation of coupled nuclear heat energy deposition and transfer in the HEU pin-type fuel assembly, as well as for steady-state safety analysis of LEU core of the GHARR-1 reactor [14,15]. e verification and validation of the PLTEMP/ANL code were also carried out by comparison with experimental data and calculation results for thermohydraulic analysis of research reactors using MTR fuel [16].…”

Section: Introductionmentioning

confidence: 99%

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

Nguyen

Huynh

et al. 2021

Science and Technology of Nuclear Installations

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This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

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Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

Cited by 4 publications

References 0 publications

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

A User’s Guide to the PLTEMP/ANL Code (V.4.3)

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

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