Reactivity Insertion Research Articles

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

Abstract To support the utilization of the power ramp test facility (PRTF) at the reactor Serba Guna GA Siwabessy (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 program for the analysis of reactor transient (PARET) and Monte Carlo N-particle version 6 (MCNP6) codes were integrated for thermal-hydraulic and neutronic analysis under the reactivity insertion phenomenon. The reactivity insertion and linear heat rate (LHR) were calculated by MCNP6 to confirm that the maximum values are well below the safety limits. Also, the code computed the reactor kinetic parameters to provide inputs for the PARET 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 safety analysis indicated that both maximum reactivity insertion and LHR values did not exceed the safety limits. The target induces maximum reactivity insertion of 0.023%dk/k and produces the maximum LHR of 137.53 W/cm and 275.06 W/cm; when the reactor operates at 15 MW and 30 MW, respectively. The PARET simulation results showed that the reactor power oscillated due to the target movement. Therefore, the recommended velocities of the target are below 0.17 m/s and 0.08 m/s for reactor power of 15 MW and 30 MW, respectively, to avoid the reactor scram, and no safety limits are exceeded.

Neutronic performance of the hybrid-low power research reactor for education and neutron applications

Korea Atomic Energy Research Institute (KAERI) has designed a second-generation Hybrid-Low Power Research Reactor (H-LPRR), which can be used as a critical assembly and conventional research reactor simultaneously. Particularly, the fuel assemblies and the reflectors of the H-LPRR are designed to stand alone without any supporting equipment and to replace each other. Its eight irradiation holes are located on the edge of the core and are applicable to neutron activation analysis and radioisotope production. In this study, some of the important nuclear characteristics are provided such as the control absorber rod worth, shutdown margin, safety reactivity factor, which are calculated by using the MCNP6.1 code with the ENDF/B-VII.0 library. Following a reactivity insertion accident, the power trend as a function of the time was also analyzed using the RELAP5/MOD3.3 code. The main safety parameters of the H-LPRR and MNSR were compared to quantitatively evaluate the neutronics performance of the H-LPRR. From these results, it can be confirmed that the H-LPRR possesses high performance in terms of nuclear design and safety analysis, and it can be a suitable option for countries considering replacement of existing and aging low power research reactors.

Transformational challenge reactor analysis to inform preconceptual core design decisions: Sensitivity study of transient analysis in a hydride-moderated microreactor

The Transformational Challenge Reactor (TCR) program aims to demonstrate a revolutionary design approach enabled by advanced manufacturing and data analytics in the design of nuclear reactors. This article discusses scoping analyses of preconceptual designs to inform TCR design decisions and the evaluation of sensitivities and uncertainties on postulated transient scenarios. The applicability of the systems codes TRACE and RELAP5-3D to TCR transient analysis are examined, and RELAP5-3D models are used to examine the transient response of two candidate core designs at multiple power levels. Then, the uncertainty quantification code RAVEN is used to quantify the effect of several design parameters on reactivity-initiated accident (RIA) progression at hot zero power (HZP) and hot full power (HFP) as well as to assess the impact of uncertainties in transient parameters for the pressurized loss of forced cooling (PLOFC).When results were compared, TRACE and RELAP5-3D showed good agreement in their ability to predict system behavior, but RELAP5-3D calculations were closer to analytical predictions for the RIA. Models for a PLOFC accident in two designs (a UO2 and TRISO core) at multiple power levels showed greater temperature margins for the TRISO core at all power levels. Using this information, along with other scoping analyses and constraints, the TCR design team selected a power level of 3 MWth and a TRISO-based core design. For this design RAVEN was applied to vary RIA parameters in RELAP5-3D models at HZP and HFP to understand the effect on figures of merit such as peak power, fuel and coolant temperature, and energy deposition. This sensitivity study found that the inserted reactivity worth was the most important parameter controlling all figures of merit, but for insertion up to 1.5$ no failure of TCR fuel is anticipated. For constant reactivity insertions, the magnitude of the fuel temperature coefficient was found to have the greatest effect on all figures of merit under most circumstances. These results not only demonstrate the anticipated robust safety of the proposed TCR fuel form but also provide a reference for future metal-hydride moderated systems to understand RIA behavior. In the PLOFC, the impact of heat transfer enhancement due to wavy flow channel effects dominated the variance in peak temperatures, and variations in heat exchanger elevation provided the greatest control on natural circulation flow rate. No fuel particle failure is anticipated in the PLOFC.

ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

Nusselt correlation development in unsteady laminar gas flows for CABRI multiphysic simulations with CATHARE2

CABRI is an experimental pulse reactor, which aims at studying and thus at better understanding RIA (Reactivity Initiated Accident) effects on nuclear fuels. The restart of the CABRI reactor in 2015 offers the opportunity to validate tools involving multiphysic calculation schemes on reactivity insertions (RI) transients at a system scale. The reactivity insertion in CABRI is mastered by the depressurization of a neutron absorber (3He), contained into transient rods. This enables the realization of various types of transients. As this depressurization highly influences the reactivity insertion kinetics, it is essential to accurately simulate the 3He density evolution which depends on the heat exchanges inside the circuit. The challenge is thus to manage the simulation of the various CABRI complex transients. Previous work allowed to model with CATHARE2 the whole reactor with its specific phenomena as well as phenomena generally involved in RIA, for instance strong neutronic feedback effects, effect of the pellet-clad interaction on the gap conductance and transient heat exchanges between fuel rods and water. These studies also revealed that some CABRI transients were still difficult to model with CATHARE2 and this paper shows the necessity to catch the transient heat exchanges in the transients rods for the simulation of the power pulse in CABRI reactor. In order to achieve this objective, a correlation for the Nusselt evolution in an unsteady compressible laminar 3He flow, inside a closed tube, has been realized from numerical resolution of an analytical model and implemented in the scientific calculation tool CATHARE2. This paper describes this analytical model and its numerical resolution. After having been validated on results of the literature in steady flow inside open tube, the correlation established for the evolution of the Nusselt in a transient rods’ tube is given. Finally, Best-Estimate simulation results of CATHARE2 are compared with experimental data obtained on CABRI commission tests: core power and 3He pressure evolution.

Development of a simple model for estimating the design limit of core void reactivity to prevent re-criticality of MOX-fueled cores in liquid metal-cooled fast reactors

The mixed oxide (MOX)-fueled core for liquid metal-cooled fast reactors has such high core performance and favorable neutron economy as to achieve an average fuel burnup of over 90 GWD/t with high power density. However, traditional MOX-fueled cores which are the ones without specific design measures to reduce the positive void reactivity worth has the potential to exceed prompt criticality during hypothetical core disruptive accidents (CDAs). Should the prompt criticality be exceeded, the core materials would heat up until their temperatures would exceed their boiling points. As a result, a release of significant mechanical energy cannot be ruled out, imposing serious damage on the reactor vessel. The released mechanical energy, termed “energetics,” is dominated by a rapid insertion of positive reactivity due to a fuel–coolant interaction (FCI) during the initiating phase of an unprotected loss-of-flow (ULOF) event. Considering the mechanism of energetics generation, a simple model was developed to estimate the design limit of positive component of core void reactivity against a ULOF event for traditional MOX-fueled cores. This simple model is constructed to evaluate the design limit based on the relation between the allowable positive reactivity insertion rate defined by the core characteristics and that assumed by the effect of FCI phenomena. The validity of the model is discussed by comparison with SAS4A which is the CDA analysis code devoted to calculate the initiating phase.

Severe accident studies on the efficiency of mitigation devices in a SFR core with SIMMER code

Sodium-cooled Fast Reactors (SFR) are investigated as future Generation IV reactor concepts. In SFR, the core configuration is not the most reactive during the nominal reactor operation, its geometry change or coolant voiding can induce a reactivity insertion. Within these considerations, mitigation devices should be implemented into the core in order to limit the thermal energy released into the fuel during a severe accident and thereby the possibility to induce mechanical loadings of the reactor structure when vaporized materials expand. Complementary safety devices called Transfer Tubes (TT) are studied in a French concept as an effective mitigation measure of severe accidents to reach a final reactor safe state, even in case of a failure of all shutdown systems including the passive shutdown safety rods. More precisely the TT, with their upper part located in and above the core, cross the core support structure and are linked to the core catcher zone situated in the main vessel bottom. Their purpose is to extract fissile materials from the core zone and secondly to favor the transfer of the molten fuel from the core region to the core catcher in case of severe core damage. The fuel discharge out of the core zone is necessary in order to lower the core reactivity, the relocation into the core catcher is necessary to enable the stabilization, the sub-criticality and the fuel cooling. An extensive work on severe accident calculations and phenomena identification related to fuel discharge through TT was performed. These studies represent the outcome of the work performed in 2014–2019, performed with collaboration of Japanese JAEA and MFBR and Framatome in the framework of the SFR project carried out by CEA. Calculations of fuel discharge were carried out with the mechanistic calculation code SIMMER and the main results are presented in this paper. Firstly, the whole unprotected loss of flow (ULOF) accident sequence was calculated with the SIMMER code. From these calculations, the power and flow evolution were set as boundary conditions for a more detailed model with six fuel assemblies around TT. This model was used to focus on key phenomena and to check impact of design parameter as local bottom restriction inside TT on fuel discharge way.

Scalable modular dynamic molten salt reactor system model with decay heat

Nodal dynamic models allow rapid reduced-order simulations of complex systems. This paper presents a publicly available system dynamic model of a molten salt reactor system with a topology of the Molten Salt Reactor Experiment (MSRE), in MATLAB/Simulink, building upon a model previously published and validated with MSRE data. It adds scaling of nominal power, dynamic representation of decay heat generation, a generic decay heat removal system, improved documentation, modularity, and test cases. One hour of system transient runs in less than 10 min on a laptop. This model is intended for first order engineering calculations, developing insights into system transient behavior, equipment sizing, parameter sensitivity studies. The model contains an easy to use toolkit which can be customized and expanded to describe any molten salt reactor system. Scenarios involving off-normal transients are presented. Relation between shutdown reactivity insertions, decay heat removal, core recriticality, and safe shutdown conditions are discussed.

Supercritical Transient Analysis for Ramp Reactivity Insertion Using Multiregion Integral Kinetics Code

To proceed with the decommissioning of the Fukushima Daiichi Nuclear Power Station, analyses of unexpected fuel debris criticality accidents are needed. Supercritical transient analyses have been conducted for fuel debris using the Multiregion Integral Kinetic (MIK) code, which can take the space dependence of fuel debris into account. In those analyses, reactivity is assumed as stepwise insertion because the MIK code does not include delayed neutron effects, which might be negligible. However, reactivity insertion may not always be stepwise. Therefore, it is important to clarify an applicable range of the MIK code for nonstepwise insertion, such as ramp reactivity insertion. To show that kinetics codes without delayed neutron effects could be applied for a supercritical transient induced by ramp reactivity insertion, we established a method to clarify its applicable range. An analysis using the point reactor kinetics model was introduced as a pre-analysis to clarify this range in the case of ramp reactivity insertion in terms of the contribution of delayed neutrons. We applied the methodology to a simple cylindrical fuel debris system and successfully demonstrated a supercritical transient analysis for ramp reactivity insertion using the MIK code.

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.

3-D TRANSIENT COUPLED SIMULATION OF SUPERPHENIX WITH PARCS/ATHLET

Most safety criteria for Sodium cooled Fast Reactors (SFR) are local core parameters. Thus, application of 3-D neutron kinetic and thermal-hydraulic coupled codes including detailed modelling of core expansion effects is mandatory for best estimate evaluations of safety margins. A recently published benchmark based on measurements performed at the Superph´enix (SPX) reactor offers the opportunity to validate codes and methods for SFR safety assessment. In this paper, the SPX core is modelled in ATHLET and PARCS. Explicit models for axial and radial core expansion effects for 3-D coupled calculations were recently implemented in PARCS. In ATHLET, axial thermal expansion of structures (strongback, vessel, control rod drive line) influencing the relative position of the control rod in the active core is modelled. The newly implemented models are tested on a transient initiated by a reactivity insertion of -50 pcm. A point kinetic simulation is also performed to compare with the 3D solution. Both ATHLET-point kinetic model and ATHLET-PARCS simulations deliver similar power responses during the transient but with an offset. By analysing the different feedbacks in the point kinetic model, it can be concluded that models of expansion of the different structures are well implemented. In the future further analysis of different transients of the benchmark are planned.

INFLUENCE OF FUEL FLOW RATE VARIATION ON MOLTEN SALT REACTOR PERFORMANCE

Molten Salt Reactors are Gen-IV reactors that use liquid fuel. Fluid fuel allows continuous removal of fission gases as well as batch fuel reprocessing. With these control mechanisms the system can be sustained within the desired operating temperature range and required power output. These methods rely on the presence of a chemical processing plant on-site that adds complexity. This also creates a risk of processing plant unavailability due to faults, emergency downtime or maintenance. The work considers variation of fuel salt flow rate in Molten Salt Reactors as a means of controlling reactor operation without using reprocessing. The analysis is performed using the Molten Salt Fast Reactor as an example. An extended version of the SERPENT Monte-Carlo transport code coupled with OpenFOAM generic platform were used for capturing delayed neutron drift, decay heat, gaseous fission product removal, calculating fuel salt velocity vectors and the fuel temperature distribution. The two models were coupled via a script that accounted for reactivity insertion between time steps and the changes caused in the fission power. Results confirm that, while operating at constant power, the difference between fuel inlet and outlet temperatures increase as the flow rate decreases. Burnup analysis has shown that while the average fuel temperature continues to reduce with time, the difference between inlet and outlet temperatures can be controlled by varying the flow rate while maintaining constant power. Finally, the variation in the fuel flow rate has been shown to extend the reactor operating time with no insertion of additional fissile inventory.

Transient calculation on TRIGA 2000 of plate-type fuel element using RELAP5, EUREKA2/RR, and MTR-DYN codes

The TRIGA 2000 reactor Bandung is proposed to convert from rod-type fuel to plate-type fuel because there are no manufacturers in the world that still produce rod-type fuel. Calculating the safety parameters and their reliability related to reactor operation has become an important thing to do. The objective of this study was to perform the reactor transient calculation of TRIGA 2000 that uses plate-type fuel. The reactor core modification is based on the capability of the existing primary coolant system, without changing its flow rates. Thermal-hydraulics calculations related to reactivity insertion accident and loss of flow accident were analyzed using EUREKA2/RR, MTR-DYN, and RELAP5 codes. The chosen loss of flow accident scenario is a decrease in flow caused by a sudden stop of the pump power supply, while reactivity insertion accident is conducted by the withdrawal of control rods with maximum reactivity insertion of 32.33 ? 10?5 s?1. The calculated parameters are reactivity, reactor power, and temperature of the coolant, cladding, and fuel in the hottest channel position. In general, from a safety point of view, those computer codes were capable of performing the transient calculations with appropriate results. Based on the calculation results, during the transient condition of both reactivity insertion accident and loss of flow accident scenario, the reactor operation safety parameters do not exceed the allowable safety limit.

Development of fuel behavior analysis code for mechanical fuel cladding failure during reactivity insertion event in PWR

Development of fuel behavior analysis code for mechanical fuel cladding failure during reactivity insertion event in PWR

BENCHMARK EVALUATION OF REACTIVITY EFFECTS AND REACTIVITY COEFFICIENTS IN THE MOLTEN SALT REACTOR EXPERIMENT

A set of benchmarks based on the experimental data from the Molten Salt Reactor Experiment (MSRE) is being compiled as part of International Reactor Physics Experiments Evaluation Reactor Physics Experiments Evaluation Project (IRPhEP). The initial benchmark that will be available in the 2019 edition of the IRPhEP handbook covers the first zero-power criticality experiment. Follow up benchmarks are under development based on the series of control rod calibration experiments performed at the MSRE, which consisted in progressive addition of a small amount (85g) of 235U in the salt followed by the insertion of the control rods acts to compensate for the excess reactivity insertion. Multiple reactivity effects and coefficients measurements are included in the benchmark: differential worth of a control rod, reactivity equivalent of 235U addition, control rod bank worth, reactivity effect of fuel circulation, isothermal temperature coefficient and fuel temperature coefficient. An uncertainty of 2% is attributed to the reported reactivity measurements from experimenters and it was believed that the uncertainty of reactor period measurement contributed the most of the experimental uncertainty. An additional 2% uncertainty was added to all reactivity measurements to represent the uncertainty for the correction factor applied to pull all the measurements on the same uranium concentration and this uncertainty was reasonably inferred by evaluating this factor on the MSRE benchmark model. The calculated reactivity equivalent of 235U additions (0.2228±0.0014, represented as the change of reactivity over the relative change of 235U mass in loop) matches well with the experiment value (0.223±0.006). Most of other calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients fall within one standard deviation from the experimental values as well.

Predication of the SMR Critical Core Performance Under Zero Power

In Nuclear Power plants, Reactivity Induced Accidents can lead to sever accidents. Rod Ejection Accidents are part of Reactivity Induced Accidents that induced through driven by reactivity insertion due to many failures. Thus, safety analysis of core behaviour under many external rod reactivities in Nuclear power plants are mandatory by regulators or safety authorities. In this research, a new dynamic model is proposed for core safety analysis under Rod Ejection Accidents. Thermal Power and other core parameters predictions are the most important goals for any reactor operation policy, during all periods and specifically at zero power to avoid severe accidents. The proposed model involves of a point kinetics explanation of neutronics combined with thermal hydraulic dynamics in the reactor core to predict its variation of parameters during transients using MATLAB environment. The proposed model is validated through comparing with the transient dynamic responses obtained through previous research for a chosen design of NuScale small modular reactor. In addition, the proposed model is verified through determining the dynamic reactor responses of Rod Ejection Accidents at hot zero power with many perturbations of different control rod ejection. The Performed safety analysis results of validation and the verification demonstrate that, the proposed model represents the reactor core behavior during the rod ejection transients with good prediction of thermal power of core peaks. Moreover, it allowed large explorations of core safety parameters and predicting the performance of its rector core during Rod Ejection Accidents under critical Hot zero power.

A best estimate plus uncertainty safety analysis framework based on RELAP5-3D and RAVEN platform for research reactor transient analyses

This paper develops a best estimate plus uncertainty (BEPU) framework for research reactor transient safety analysis. The BEPU framework is developed based on the system level reactor safety analysis code RELAP5-3D and the data analysis platform RAVEN developed by Idaho National Laboratory. Within the framework, a sensitivity analysis procedure is first conducted to identify the contributions and ranks of individual input parameters to the user-defined figures-of-merit (FOMs) associated with specific transient phenomena. An uncertainty analysis procedure is then performed to quantify the uncertainties of the FOMs resulting from the uncertainties of the input parameters. Many useful outcomes can be realized through the BEPU analysis. Specifically, the sensitivity information obtained from the sensitivity analysis will provide insights about the influence of each different input parameter on FOMs. The uncertainty information obtained from the uncertainty analysis will imply the range of response deviations caused by the propagation of errors existing in various input components.As a case study for research reactors, the developed BEPU framework was employed to perform design-basis accident (DBA) analysis for one conceptual research reactor design proposed at the National Institute of Standards and Technology (NIST). Two hypothetical DBA scenarios, namely the reactivity insertion accident (RIA) and the loss of flow accident (LOFA), were modeled and analyzed through the BEPU framework. To demonstrate the value of the BEPU framework, the BEPU analysis results were compared to that obtained from the conventional transient safety analysis procedure, which was conducted by using commonly used transient safety analysis codes including RELAP5-3D and PARET. The comparison shows that the BEPU analysis is capable of providing additional sensitivity and uncertainty information that help confirm safety margins of the NIST conceptual research reactor during both RIA and LOFA situations, which justifies the advantages and benefits of the BEPU safety analysis framework developed in this work.

Simplified criteria for a comparison of the accidental behaviour of Gen IV nuclear reactors and of PWRS

A cross-comparison of four Generation IV reactor concepts and a PWR of 2nd generation is presented in this paper. 4 Gen IV reactor concepts are considered and are briefly presented in the first part of the paper: SFR of 1500 MWth, GFR of 2400 MWth, MSR of 3000 MWth and VHTR of 600 MWth. In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria aims at assessing the main physical thresholds to exceed in order to have a core degradation: phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assessed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is evaluated. A third set of criteria deals with core features and aims at assessing the possible reactivity insertion that withstands each concept up to core melting (or boiling for the MSR) and the associated expected power peaks in case of coolant voiding/depressurization and in case of fissile material compaction. Finally, a last criterion set deals with the assessment of the possibility to jeopardize physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant/fuel and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concepts investigated in terms of severe accident prevention and mitigation.

Neutronic design and dynamic analysis of a 450 MWth graphite molten salt reactor core

Molten salt reactor (MSR) has good economic potential, inherent safety and nonproliferation characteristics. In this paper, the 450 MWth graphite molten salt reactor is taken as the research object. First, the neutronic design of a 450 MWth graphite molten salt reactor is introduced in terms of the core material, conceptual design, static physical characteristics and core design parameters. Then, the model of the MSR core is established by combining the neutron dynamics model, the thermal model and the auxiliary calculation. Finally, the transient characteristics of the 450 MWth MSR core under the conditions of step reactivity insertion and main pump coast-down are simulated and analyzed. The results show that under the aforementioned conditions, the core power can be stable in the controllable level range, and the core temperature is in a safe range.

The influence of neutron reflector with special properties on fast reactor kinetics

The paper substantiates the possibility of increasing the safety of fast reactors by surrounding the reactor core with a neutron reflector with special properties. The particular case of using lead isotope as a neutron reflector has been considered in the lead-cooled fast reactor in an exemplary manner. An asymptotic point kinetics model, which takes into account neutrons returning from the reflector to the reactor core and participating in the development of fission chain reaction, was proposed for assessment of the emerging effects. The in-hour equation for fast reactors, surrounded by a neutron reflector with special properties, has been developed. It was revealed that the neutron lifetime can be substantially increased. Therefore, the contribution of the neutron reflector would mitigate the consequences of accidents when inserted reactivity exceeds the effective fraction of delayed neutron. In addition, the maximum rates of power increase have been calculated for different cases of reactivity insertion. The obtained results have considerable practical value for scientific and advanced studies of reactor accidents.