• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1156-1163
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• 2020

Fuel temperature coefficient (FTC) of PuO₂+ZrO₂ (ROX) fueled LWR cell is analyzed neutronically with reactor- and weapons-grade plutonium fuels in comparison with a U-free PuO₂+ThO₂ (TOX), and a conventional MOX fuel cells. The FTC value of a ROX fueled LWR is smaller compared to a TOX or a MOX fueled LWRs and becomes extremely positive especially, at EOL. This is because when fuel temperature is increased, thermal neutron spectrum is shifted to harder, which is extreme at EOL in ROX fuel than that in TOX and MOX fuels. Consequently at EOL, ²³⁹Pu and ²⁴¹Pu contributes to positive fuel temperature reactivity (FTR) in ROX fuel, while they have negative contribution in TOX and MOX fuels. The FTC problem of ROX fuel is mitigated by additive ThO₂, UO₂ or Er₂O₃. In ROX-additive fuel, the atomic density of fissile Pu becomes more than additive free ROX fuel especially at EOL, which is the main cause to improve the FTC problem. The density of fissile Pu is more effective to decrease the thermal spectrum shifts with increase the fuel temperature than additive ThO₂, UO₂ or Er₂O₃ in ROX fuel.

• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.488-499
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• 1997

The measurements of the moderator temperature coefficient (MTC) are performed to demonstrate that the calculational model produces results that are consistent with the measurements. Since negative MTC is also a technical specification value that may limit the cycle length, it is important to measure it as accurately as possible. In this report, preferred choice of test method depending on the time in cycle, best power indication and temperature definition in MTC calculation were determined based on the MTC test results taken during initial startup testing and at 2/3 cycle burnup in the Yonggwang nuclear power plant. The results show that the ratio and rodded methods provided good agreement with the predictions during initial startup testing. However, near end-of-cycle the depletion method gives better results, and so is suggested to be used in the MTC measurements at 2/3 cycle burnup. The use of primary Delta T power as a power indicator in the MTC calculations is highly advisable since it responds with good consistent results very quickly to changes unlike secondary calorimetric power. For the appropriate temperature definitions used in the MTC calculations, it is considered that the arithmetic average temperature measured simply by inlet and outlet thermocouples is preferred. Although volumetric average temperature provides better results, the improvement is not sufficient to compensate for the simplicity of calculations by arithmetic average temperature.

• 한국연소학회:학술대회논문집
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• 2014.11a
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• pp.179-182
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• 2014

A comparative DNS study of the ignition characteristics of dual-fueled reactivity controlled compression ignition (RCCI) and stratification charge compression ignition (SCCI) is investigated using a 116-species reduced primary reference fuel (PRF) mechanism. In the RCCI combustion, two PRF fuels (n-heptane and iso-octane) with opposite autoignition characteristics are separatedly supplied and in-cylinder blended such that spatial variations in fuel reactivity, fuel concentration and temperature are achieved. In the SCCI combustion, however, just a single fuel (PRF50) is used such that only fuel concentration and temperature inhomoginieties are obtained. Because three factors, rather than only two as in SCCI combustion, govern the overall RCCI combustion, combustion timing and combustion duration or heat release rate of RCCI combustion are flexibly and effectively controlled. It is found that the overall RCCI combustion occurs much earlier and its combustion duration is longer compared to SCC combustionI. Moreover, the negative temperature coefficient (NTC) has a positive effect on enhancing RCCI combustion by inducing a shorter combustion timing and a longer combustion duration as a result of the occurrence of a predominant low-speed deflagration-combustion mode.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.639-647
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• 2018

In this study, the effects of selecting water/silver nanofluid as both a coolant and a reactivity controller during the first operating cycle of a light water nuclear reactor are investigated. To achieve this, coupled neutronic-thermo-hydraulic analysis is employed to simulate the reactor core. A detailed VVER1000/446 reactor core is modeled in monte carlo code (MCNP), and the model is verified using the porous media approach. Results show that the maximum required level of silver nanoparticles is 1.3 Vol.% at the beginning of the cycle; this value drops to zero at the end of cycle. Due to substitution of water/boric acid with water/Ag nanofluid, reactor operation time at maximum power extends to 357.3 days, and the energy generation increases by about 27.3%. The higher negative coolant temperature coefficient of reactivity in the presence of nanofluid in comparison with the water/boric acid indicates that the reactor is inherently safer. Considering the safety margins in the presence of the nanofluid, minimum departure from nucleate boiling ratio is calculated to be 2.16 (recommendation is 1.75).

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3897-3908
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• 2022

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on "Neutronics Benchmark of CEFR Start-Up Tests" offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3516-3525
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• 2022

The high capture cross-section (𝜎_c) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high 𝜎_c as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd₂O₃ and Er₂O₃ respectively. Models comprise nuclear fuel with a homogeneous mixture of Er₂O₃, AmO₂, SmO₂, and HfO₂ with Gd₂O₃ as well as the coating of PaO₂ and ZrB₂ on the Gd₂O₃ pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er₂O₃ and SmO₂ showed expected results namely higher values of these parameters than the reference design of Gd₂O₃ at the early burnup period. The highest value was found for the model of PaO₂ and Gd₂O₃ followed by ZrB₂ and HfO₂. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd₂O₃ and absorber coating thickness.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.801-805
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• 2018

In this study, an underwater radiation detector was built using a GAGG(Ce) scintillator and silicon photomultiplier to establish an underwater radiation exposure monitoring system. The GAGG(Ce) scintillator is suitable for small radiation detectors as it strongly absorbs gamma rays and has a high light emission rate with no deliquescent properties. Additionally, the silicon photomultiplier is a light sensor with characteristics such as small size and low applied voltage. Further, a program and mobile app were developed to monitor the radiation coefficient values generated from the detector. According to the results of the evaluation of the characteristics of the underwater radiation monitoring system, when tested for its responsiveness to radiation intensity and reactivity, the system exhibited a coefficient of determination of at least 0.99 with respect to the radiation source distance. Additionally, when tested for its underwater environmental temperature dependence, the monitoring system exhibited an increase in the count rate up to a certain temperature because of the increasing dark current and a decrease in the count rate because of decreasing overvoltage. Extended studies based on the results of this study are expected to greatly contribute to immediate and continuing evaluation of the degree of radioactive contamination in underwater environments.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.345-355
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• 2019

Use of the reconstructed fuel assemblies from partially burnt nuclear fuel pins is analyzed. This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin ${\eta}$ values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with ${\eta}$ > 1:0 are used to reconstruct to-be-reused fuel assemblies. These reconstructed fuel assemblies are burnt during the cycle 3, 4, 5 and 6 of a 1000 MW PWR core by replacing fresh, once burnt and twice burnt fuel assemblies of the reference core configurations. It is concluded that using reconstructed fuel assemblies for the fresh fuel affect dearly on the cycle length (>50 EFPD) when more than 16 fresh fuel assemblies are replaced. However, this loss is less than 20 days if the number of fresh fuel assemblies is less than eight. For the case of replacing twice burned fuel, cycle length could be increased slightly (10 days or so) provided burnt fuel pins from other reactors were also available. Reactor safety parameters, like axial off set (< ${\pm}10%$), Doppler temperature coefficient (<0), moderator temperature coefficient at HFP (<0) are always satisfied. Though, 2D and 3D pin peaking factors are satisfied (<1:55) and (<2:52) respectively, for the cases using eight or less reconstructed fuel assemblies only.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.