• 전기의세계
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• 제15권5호
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3106-3116
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• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.648-653
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• 2018

Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR) design, only control rods are available to control such rapid core transient. The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model. The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2048-2054
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• 2022

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.573-580
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• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1165-1203
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• 2024

In the paper we validate theoretical models of the pulse against experimental data from the Jozef Stefan Institute TRIGA Mark II research reactor. Data from all pulse experiments since 1991 have been collected, analysed and are publicly available. This paper summarizes the validation study, which is focused on the comparison between experimental values, theoretical predictions (Fuchs-Hansen and Nordheim-Fuchs models) and calculation using computational program Improved Pulse Model. The results show that the theoretical models predicts higher maximum power but lower total released energy, full width at half maximum and the time when the maximum power is reached is shorter, compared to Improved Pulse Model. We evaluate the uncertainties in pulse physical parameters (maximum power, total released energy and full width at half maximum) due to uncertainties in reactor physical parameters (inserted reactivity, delayed neutron fraction, prompt neutron lifetime and effective temperature reactivity coefficient of fuel). It is found that taking into account overestimated correlation of reactor physical parameters does not significantly affect the estimated uncertainties of pulse physical parameters. The relative uncertainties of pulse physical parameters decrease with increasing inserted reactivity. If all reactor physical parameters feature an uncorrelated uncertainty of 10 % the estimated total uncertainty in peak pulse power at 3 $ inserted reactivity is 59 %, where significant contributions come from uncertainties in prompt neutron lifetime and effective temperature reactivity coefficient of fuel. In addition we analyse contribution of two physical mechanisms (Doppler broadening of resonances and neutron spectrum shift) that contribute to the temperature reactivity coefficient of fuel. The Doppler effect contributes around 30 %-15 % while the rest is due to the thermal spectrum hardening for a temperature range between 300 K and 800 K.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1715-1720
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• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.