Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

Boulaich, Y.; Nacir, B.; Bardouni, T. El; Boukhal, H.; Chakir, E.; Bakkari, B. El; Younoussi, C. El

doi:10.1016/j.nucengdes.2014.12.037

Cited by 3 publications

(1 citation statement)

References 4 publications

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“…Jiang et al 13 conducted the transient analysis of Thorium‐based molten salt reactor with solid fuel (TMSR‐SF) to evaluate the safety performance of reactor system based on modified RELAP5/MOD4.0.Y. Boulaich et al 14 established a three‐dimensional model for the Moroccan 2 MW TRIGA MARK II reactor. Based on this thermal‐hydraulic model, a safety analysis under the reactivity insertion phenomenon was carried out and the dependence on time of the main thermal‐hydraulic parameters was calculated.…”

Section: Introductionmentioning

confidence: 99%

Transient thermal‐hydraulic analysis of heat pipe cooled passive residual heat removal system of molten salt reactor

et al. 2020

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Summary As one of the representatives of fourth generation nuclear reactors, molten salt reactor technology has become the research focus worldwide. Higher passive safety requirements have been put forward for the nuclear reactor systems. In this paper, a conceptual design of a passive residual heat removal system for the molten salt reactor is presented and the mathematical and physical models are established. Based on FORTRAN programming language, the passive residual heat removal system analysis code is developed and validated. The code is used to analyze the safety of molten salt reactor under accident conditions, including unprotected reactivity introduction accident, unprotected overcooling, unprotected loss of flow, opening failure of the frozen valve and the startup accident of partial heat pipes. The results show that the passive residual heat removal system of molten salt reactor can discharge fuel salt decay heat effectively within 200 hours and ensure that the molten salt reactor to operate safely under all accident conditions. Above all, the system has good passive safety performance. The proposed passive residual heat removal system can satisfy the safety requirement of molten salt reactor. The feasibility of the conceptual design of the heat pipe cooled passive residual heat removal system is theoretically verified.

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Section: Introductionmentioning

confidence: 99%

Transient thermal‐hydraulic analysis of heat pipe cooled passive residual heat removal system of molten salt reactor

et al. 2020

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Safety Analysis of Indonesia's Multipurpose Reactor During Irradiation of Natural UO2 Pin Target at Power Ramp Test Facility

Fadli

Khakim

Suharyana

et al. 2021

Journal of Nuclear Engineering and Radiation Science

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In order to support the utilization of Power Ramp Test Facility (PRTF) at RSG-GAS reactor, the investigation of safety parameters both on the target and the reactor during the irradiation of natural UO2 has been conducted. The PARET/ANL and MCNP6 codes are integrated for thermal-hydraulic and neutronic analysis under the reactivity insertion phenomenon. In the MCNP6 code, the reactivity insertion and Linear Heat Rate (LHR) are analyzed to confirm that the maximum value is still below the safety limits. In addition, the code computes kinetic parameters to provide inputs for the PARET/ANL code to evaluate the reactor response due to the target movement at the PRTF. The target moves from the farthest position to the nearest one traveling 440 mm, and then moves back to the original position. The result of the irradiation target simulation is satisfying, both maximum reactivity insertion and LHR values do not exceed the safety limit allowed. The pin target induces maximum reactivity insertion of 0.023%dk/k and produces the maximum LHR of 137.27 watt/cm and 274.54 watt/cm; when the reactor operates at 15 MW and 30 MW, respectively. In addition, the PARET/ANL simulation indicates that the reactor power oscillates due to the target movement. In order to avoid the reactor scram, the recommended pin target maximum velocities are 170 mm/s and 80 mm/s for reactor power of 15 MW and 30 MW, respectively.

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Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes

et al. 2018

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Innovative Research Reactor 50 MW (RRI-50) is a conceptual design of an Indonesia's high-power innovative research reactor, which is designed to produce a maximum neutron flux as high as 1.0 × 1015 n/cm2 s and thermal neutron flux of 5.0 × 1014 n/cm2 s. In this study, the design basis accident slow reactivity insertion is carried out. Reactivity insertion accident (RIA) caused by the inadvertent withdrawal of all fuel rod at normal rate is simulated by using PARET-ANL transient code. In the simulation accident is assumed to occur when the reactor is in operation at high power and at low power level. During transient, the fuel rods are assumed to be in the most effective positions with insertion rate of 0.049 $/s. This reactivity worth will give effect only if the fuel rod interlock system is failed to function properly and the operator action does not follow the standard operation procedure. The analysis results show that the reactor experiences scram when it reaches its protected power at high power level of 110 % (55 MW) at period of 5 s and at low power level of 3 MW at period of 120 s, and there is sufficient safety margin for the anticipated RIA caused by reactivity feedback.

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Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

Cited by 3 publications

References 4 publications

Transient thermal‐hydraulic analysis of heat pipe cooled passive residual heat removal system of molten salt reactor

Transient thermal‐hydraulic analysis of heat pipe cooled passive residual heat removal system of molten salt reactor

Safety Analysis of Indonesia's Multipurpose Reactor During Irradiation of Natural UO2 Pin Target at Power Ramp Test Facility

Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes

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