Reactivity Insertion Research Articles

Transient analysis of tritium transport characteristics in fluoride-salt-cooled high-temperature reactor

Tritium control is a challenging issue for the commercialization of Fluoride-salt-cooled High-temperature Reactor (FHR). In this paper, the tritium behaviors in a FHR system are briefly introduced. A tritium transport characteristics analysis code (the updated TAPAS code) is developed and validated to research the dynamic responses of tritium behaviors during reactor accidents. Two typical accidents for the FHR are simulated, including unprotected reactivity insertion accident (URIA) and unprotected overcooling (UOC) accident. Numerical results show that tritium transport characteristics are strongly coupled under transient conditions; they are quite different from those of normal operation, ranging from tritium production and speciation, adsorption by graphite, chromium corrosion and deposition to tritium permeation through structural materials. The results are discussed in detail in comparison with our previous study. Based on current calculation models, the T2 permeation rate decreases under the URIA and UOC, therefore, extra safety facilities for tritium control under these two transient conditions are not necessary. This study may provide a valuable reference for the tritium control in the FHR.

A realistic approach for the assessment of the consequences of heterogeneous boron dilution events in pressurized water reactors

In order to compensate the excess reactivity over the fuel cycle in pressurized water reactors, boric acid is added to the reactor coolant. During certain accident conditions, the formation and the subsequent transportation towards the reactor core of a slug of coolant with reduced boron concentration can lead to a positive reactivity insertion. In this paper, a new approach dealing with such heterogeneous boron dilution scenarios is presented. This approach is based on the use of computational fluid dynamics methods for the whole reactor pressure vessel with direct coupling to a 3D neutron kinetic model of the reactor core. The application of this approach is demonstrated on a main coolant pump start scenario for hot-shutdown conditions using the coupled ANSYS CFX/DYN3D code. The calculation results show the enormous influence of the coolant mixing on the reactivity behavior of the reactor core.

Transient Reactor Test (TREAT) Facility Restart Heat Balance Calibration and Rod Worth Measurements

The Transient Reactor Test Facility has been restarted after more than 20 years in a safe standby condition. The plan to bring the reactor back into operation included a typical core characterization that was historically performed every time the core was reconfigured for a new experiment campaign. The core characterization included determining initial critical position of the control rods, a heat balance run for calibration of the nuclear instruments to enable the indication of reactor power, control rod worth measurements, and a series of three temperature-limited transients increasing in the amount of reactivity inserted as a step for the interpolation of set points for the reactor trip system and reactivity insertion limits. The heat balance and control rod worth measurements are discussed in this paper. After critical control rod position was determined, a heat balance operation was used to position the nuclear instruments for correct power indication. This was followed by control rod differential worth measurements to generate the control rod worth curves used by the automatic reactor control system for control of the reactor during transient operations. These restart evolutions are summarized here, and the results are compared to the historic measurements.

Fuel cladding damage assessment based on reactivity insertion accidents for a 10 MWth LBE cooled fast reactor

Fuel cladding integrity affects the safety and economy of nuclear reactors, especially in accident condition. In this paper, based on the reactivity insertion accident (RIA) analysis, the fuel cladding damage has been analyzed for a 10 MWth LBE cooled fast reactor developed by Institute of Nuclear Energy Safety and Technology (INEST) using RELAP5. Simulation was carried out for control rod withdrawal accident with a reactivity insertion rate of 50.19 pcm/s for 4 s, both for scram and without scram. During the control rod withdrawal accident without scram, the highest cladding temperature was 714 °C, which was below the temperature limit of cladding integrity. It is concluded that the RIA will not cause reactor safety violation due to the reactor’s inherent safety feature.

Verification and validation of the modular ray tracing MOC using the coupled forward-adjoint approach and application to C5G7 benchmark

Neutron transport adjoint calculations are useful in many reactor physics applications. Among various applications, the adjoint flux can be used in perturbation theory to prevent large calculations and computational costs when estimating a small reactivity insertion into the system. Furthermore, they can serve as validation tests for numerical schemes, since both direct and adjoint calculations for a given system should lead to the same eigenvalue, although using two different physicomathematical formulations of the transport model.The synchronized implementation of the method of characteristic (MOC) for neutron transport in forward and adjoint approaches is accomplished in this work. The result is validated using the C5G7 benchmark with comparisons of the multiplication factor and pin power values. Differences between forward and adjoint multiplication factors in the results are achieved in the order of 1.0E-6. Meanwhile, the difference between the multiplication factor and the C5G7 benchmark is in order of 1.0E-5, using an S16 level symmetric angular discretization and a track spacing of 0.01 cm.

Electric power frequency and nuclear safety - Subsynchronous resonance case study

Abstract The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink MATLAB® model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5

Nuclear power and thermionic conversion can serve as a compact, durable energy source for the space exploration and exploitation. In this paper, the modified Reactor Excursion and Leak Analysis Program5 (RELAP5) with the implement of NaK-78 eutectic alloy (78%K and 22%Na) properties and heat transfer correlations is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including the thermionic fuel elements (TFEs), reactor core, radiator, coolant loop and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, reflector, moderator and the reactivity insertion effects of control drums and safety drums are considered. The steady state condition and three severe transient accidents including reactivity insertion accident (RIA), loss of flow accident (LOFA) and loss of coolant accident (LOCA), are simulated and analyzed. The steady state calculated results agree well with the design values. During the three accidents, the moderator plays a dominant role in the positive temperature reactivity feedback. The coolant has at least 50 K temperature margin to the boiling point. The fuel and TFE components are all below their melting temperature. The progress of these accidents provide relatively sufficient time for operator's response. The calculation results prove that the reactor is a safe and reliable system.

Development of a control-oriented power plant simulator for the molten salt fast reactor

In this paper, modelling and simulation of a control-oriented plant-dynamics tool for the molten salt fast reactor (MSFR) is presented. The objective was to develop a simulation tool aimed at investigating the plant response to standard control transients, in order to support the system design finalization and the definition of control strategies. The simulator was developed employing the well tested, flexible and open-source object-oriented Modelica language. A one-dimensional modelling approach was used for thermal-hydraulics and heat transfer. Standard and validated thermal-hydraulic Modelica libraries were employed for various plant components (tubes, pumps, turbines, etc.). An effort was spent in developing a new MSR library modelling the 1D flow of a liquid nuclear fuel, including an ad-hoc neutron-kinetics model which properly takes into consideration the motion of the Delayed Neutron Precursors along the fuel circuit and the consequent reactivity insertion due to the variation of the effective delayed fractions. An analytical steady-state 2-D model of the core and the fuel circuit was developed using MATLAB in order to validate the Decay Neutron Precursors model implemented in the plant simulator. The plant simulator was then employed to investigate the plant dynamics in response to three transients (variation of fuel flow rate, intermediate flow rate and turbine gas flow rate) that are relevant to control purposes. Simulation outcomes highlight the typical reactor-follows-turbine behavior of the MSFR, and they show the small influence of fuel and intermediate flow rate on the reactor power and their strong effects on the temperatures in their respective circuits. Starting from the insights on the reactor behavior gained from the analysis of its free dynamics, the plant simulator here developed will provide a valuable tool in support to the finalization of the design phase, the definition of control strategies and the identification of controlled operational procedures for reactor startup and shutdown.

Breeding Properties Study on High-Power Thorium Molten Salt Reactor

Molten salt reactor (MSR), the only one using liquid fuel in the six types “Generation IV” reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high-power thorium molten salt reactor (TMSR) based on the two-fluid molten salt breeder reactor (MSBR) with a superior breeding performance. We explored the optimized structure to be a thorium-based molten salt breeder reactor with different core conditions and different postprocessing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 s and the core behaves safely.

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5 (RELAP5), with the implementation of sodium–potassium eutectic alloy (NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including thermionic fuel elements (TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-II system model, an electrical ground test of the fully integrated TOPAZ-II system, the V-71 unit, is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-II are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

CADOR “Core with Adding DOppleR effect” concept application to sodium fast reactors

Generation-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters.

Convex shaped FBR core with graded size pins to prevent re-criticality during Core Disruptive Accident

The core geometry of sodium-cooled fast spectrum reactor (SFR) is not generally in its most reactive configuration. Therefore, the study of postulated Core Disruptive Accident (CDA) has been recognized as one of the most important themes for SFR. There are several analyses and experiments to investigate CDA phenomena. On the basis of these knowledge, some special devices are designed to control the relocation of molten material. This study focuses on the core configuration to maximize neutron leakage effect during its geometrical deforming, instead of introduction of the devices. The authors investigate the convex shaped core design that is closed to spherical shape. It is because the oval spherical core has the least neutron leakage in all practical core geometries. The neutron leakage of the convex shaped core increases by geometry change and it causes negative compaction reactivity insertion. In this paper, the feasibility of the convex shaped core design is investigated based on the parametric survey of compaction reactivity using the neutron diffusion calculation. The convex shaped core with uniform size pins and less neutron leakage shows negative compaction reactivity but the radial power distribution is not flattened. The introduction of graded size pins enables convex core to accomplish negative compaction reactivity and leveled power, simultaneously.

Reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor

The leakage from the reactor pool back into the dry irradiation channels due to corrosion or mechanics based reason is a postulated event that could occur under operating conditions of the Dalat nuclear research reactor (DNRR), especially the channel 7-1 which has been installed more than 30 years. When it occurs, the air space in these channels will be occupied by the water, subsequently a water column will appear in fuel region. The appearance of water column considerably enhances medium of neutron moderation for its surrounding fuel assemblies. As a result, a positive reactivity is inserted in the core and this event is classified as an insertion of excess reactivity. This event needs to be addressed by analysis and assessment from safety point of view and the results of analysis are also important for updating the reactor operating procedures. This paper presents assumptions, computer models and the results of analysis for such event in the DNRR by using MCNP5 code (code for neutronics analysis) and EUREKA-2/RR code (code for transient analysis). The calculation results include value of reactivity insertion, change in power of reactor, as well as surface temperature of the hottest fuel assembly. This research contributes to updating the reactor operating procedure.

An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading

An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

Development of Monte Carlo Neutron Transport Method-Based Supercritical Transient Code with Time-Dependent Feedback Capability

A multiregion integral kinetic (MIK) code based on the integral kinetic model and a Monte Carlo neutron transport method has been developed with a new time-dependent feedback modeling capability. The current MIK code is applicable to the supercritical power transient following reactivity insertion in a fissile system of arbitrary geometry and composition, taking its feedback mechanisms into account. The new time-dependent feedback modeling capability allows a more direct and accurate treatment of complicated and nonlinear feedback mechanisms in a given system. The purpose of this study is to verify the MIK code and its time-dependent feedback modeling capability through various supercritical transient experiments conducted at the Godiva, TRACY, and SILENE facilities. Specifically, four supercritical experiments were selected and simulated using the MIK code. The various complicated feedback mechanisms—thermal expansion in Godiva, and Doppler broadening, thermal expansion, and radiolytic gas creation in TRACY and SILENE—provide a good benchmark for verifying the MIK code and its time-dependent feedback model. The obtained results show generally good, albeit occasionally poor, agreement with experimental results depending on the specific experiment. When the reasons for the poor agreement are considered, however, it may be concluded that the simulated results show promising agreement with the experiments, verifying the MIK code and its time-dependent feedback modeling capability.

Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination

Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.

Nuclear safety of the fast reactor with 208Pb-reflector at neutron flash above effective fraction of delayed neutrons

This paper aims at improving fast reactors safety through slowing-down their power runaways. The method is surrounding the core by the neutron reflector made of materials of heavy atomic weight and extremely low neutron absorption. The power runaways can be slowed down because of a long way for leakage neutrons to come back from distant layers of neutron reflector to the core. Time interval needed for leakage neutrons to come back to the core could become comparable with lifetime of delayed neutrons. So, this kind of reflector is able to transform some fraction of prompt leakage neutrons into additional group of delayed neutrons. It has been demonstrated that mean prompt neutron lifetime can be elongated roughly by three orders of magnitude with appropriate slowing-down the reactor power runaway. The most significant slowing-down the power runaways can be reached at the most dangerous reactivity insertion, i.e. at the reactivity larger than effective fraction of delayed neutrons. Thanks to the elongated mean prompt neutron lifetime and to the shortened time constant of fuel rods the negative consequences of the accidental reactivity insertion above effective fraction of delayed neutrons could be substantially weakened because, in addition to the Doppler-effect, the negative feedbacks on coolant and cladding temperatures receive enough time to actuate.

Analysis of the impacts of homogeneous minor actinides loading in low void effect sodium fast reactor cores

Minor actinides transmutation is a solution to decrease the long-term radiotoxicity of nuclear waste and limit their short-term decay heat. In the homogenous approach, minor actinides are mixed in the fuel and loaded in the core in order to turn them into fission products. This leads to a neutron spectrum hardening in the core, which has a negative impact on the core integral feedback coefficients such as the Doppler Effect or the sodium void worth. Analysis of these impacts was used in the past to establish limit on maximal minor actinides loading in a reactor core. Low-void cores (CFV in French) have been recently developed by CEA to achieve negative sodium void worth by adding axial heterogeneities in the core layout in the form of an upper sodium plenum and an inner fertile blanket. In this paper, the impacts of minor actinides loading in such a core are analyzed and a comparison is carried out with earlier homogeneous core designs. In a first time, the impacts on integral feedbacks coefficients are evaluated, and in a second time, the perturbations of the core behavior during selected representative transients are analyzed. It is shown that even if the impacts on integral coefficients are similar between the two core designs, the low-void core behavior during loss-of-flow transient is not negatively impacted by minor actinides loading. For reactivity insertion transient, the two core designs behave similarly. It can be concluded that considering heterogeneous cores, the use of integral coefficients is insufficient to characterize the impact of minor actinides loading on the core for loss-of-flow transient. However, the impacts of minor actinides loading on the Doppler integral feedback coefficient can be used as reliable estimator for the modification of the core behavior during a reactivity insertion transient.

Reactor dynamics of in-pin fuel motion in fast breeder reactors

Abstract In-pin fuel motion is a multi-phase hydrodynamic movement of molten fuel inside the fuel pins of a fast reactor during a core disruptive accident. This study focuses on the potential consequences of this phenomenon on fast breeder reactor dynamics during an unprotected transient overpower (UTOP) accident. An in-house developed solver is benchmarked with the CABRI-E9bis test to addresses the complex interplay of hydrodynamic forces. It is deduced that in slow overpower, molten fuel hydrodynamics is dominated by gravity and solidification blockage. Next, the solver is integrated with an in-house reactor dynamics code ‘PREDIS’ to simulate UTOP under the effect of in-pin fuel motion. A 500 MWe mixed oxide fuel based fast reactor is chosen for analyses. Results show that in-pin fuel motion generates a negative reactivity feedback of the order of −0.45 $. This results in a reduced peak power (182%) and stabilized state power (175%). Damage to the reactor core is contained more effectively. It is found that in a beginning of equilibrium core (BOEC), high internal pressure of the fuel pin does not permit molten fuel vaporization at high temperatures. Higher reactivity insertion rates of up to 20 pcm/s are also countered with in-pin fuel motion. From this study, it is inferred that in-pin fuel motion feedback enhances the inherent safety features of fast reactors during UTOP. It scales in magnitude with the fuel Doppler, and is therefore an important parameter for future safety analyses of oxide fuelled fast reactors.

Verification of iterative matrix solutions for multipoint kinetics equations

Multipoint kinetic equation systems have been solved numerically using matrix algebra software and 1st to 4th order implicit schemes based on single-step matrix propagators. These matrix schemes have been validated successfully on demanding point kinetics benchmarks with various prescribed reactivity insertions. Verification tests have also been performed on a simple 3-region fast reactor core model with asymmetric step or ramp reactivity insertions and cross-checked with the multipoint kinetics SACRE code developed at INL and based on existing and validated stiff ODE solver packages. Accurate results can be obtained at a limited time expense. The reason for intriguing results obtained in some transients has been elucidated and linked to the multipoint matrix coefficient interpolation method during the transient.