Rod Withdrawal Research Articles

Sensitivity analysis and fuel performance of a control rod withdrawal in the generic fluoride-salt-cooled high temperature reactor

The purpose of this study is to develop boundary conditions with respect to input parameter uncertainty for fuel performance assessment of a control rod withdrawal transient in a fluoride-salt-cooled high-temperature reactor (FHR). Sensitivity analysis was performed to determine how certain key input parameters impacted the uncertainty of the outputs needed for the boundary conditions. The generated boundary conditions can be used in fuel performance codes to obtain failure probabilities and fuel performance envelopes. To accomplish this, we utilized a KP-AGREE model of the generic fluoride-salt-cooled high-temperature reactor (gFHR) and a pebble bed benchmark developed by Kairos Power. The key boundary conditions that were developed included the total power, maximum fuel region temperature, maximum fuel kernel temperature, and maximum moderator (graphite) temperature within the core. The sensitivity analysis was run until the Sobol indices and output converged within a 95 % confidence level; it was found that the coolant inlet temperature consistently had the highest impact on the uncertainty of the key core temperature values. The maximum bound of these temperatures also did not exceed 1123.5 °C for the fuel kernel and 973.7 °C for the moderator after the transient, demonstrating that there is a large margin between the predicted fuel and moderator temperature and accepted safety limits, given a 1650 °C limit for Tristructural Isotropic (TRISO) fuel. These boundary conditions were implemented into a BISON model to demonstrate their use in fuel performance and obtain a brief failure profile of the fuel. The maximum magnitude of the hoop stress on the SiC layer was −12.02 MPa, resulting in a failure probability of 4.90·10−19. Thus, the fuel performance simulations also emphasized the robustness of the TRISO fuel design utilized in the gFHR, as the predictions reveal substantial margin of the TRISO particles relative to temperature limits during the control rod withdrawal. The limiting temperature is expected to be the reactor vessel temperature and the fuel is not expected to be challenged at all by mechanical failure of the SiC layer during these events.

Safety analysis of inadvertent control rods withdrawal of RSG-GAS research reactor for various initial powers

RSG-GAS is an Indonesian MTR-type multipurpose research reactor with a nominal power of 30 MW. Within the context of Design Basis Accidents (DBAs), numerous potential initiating events with the potential to lead to reactivity insertion accidents (RIA) have been identified. Among these, simultaneous inadvertent control rod withdrawal stands out as one of the most limiting scenarios for MTR-type research reactors. In these safety analyses, we employ the PARET code to simulate three distinct initial power conditions: 1 MW, 15 MW, and 30 MW during high power with forced convection cooling mode. The selection of 15 MW as an initial power level is based on the fact that the reactor predominantly operates at this power level. In addition, another safety evaluation was also performed during low power of 300 kW with natural convection cooling mode. Throughout these simulations, it was postulated that the first trip signal fails to initiate a reactor scram, and eventually, the second signal successfully scrams the reactor. This transient journey may activate various reactor protection systems, including but not limited to, period-based protection, floating limit value protection, overpower trip protection, and more. The analyses have confirmed that no single safety limit was exceeded.

A PWR core power control using optimized PID controller with SQP based on control rod positioning and variable coolant flow rate

To ensure safety and stability, it is optimal for nuclear power plants to operate at their nominal power level. However, to meet the needs of power systems, NPPs must also be able to adjust their output for load-following purposes. This necessitates finding an effective control algorithmto maintain stable/safe operating conditions. The current study investigates the performance of PID controllersoptimized using Sequential Quadratic Programming algorithm and applied using two control techniques: control rod positioning and variable coolant flow rate. Simulation is done using model of four-loop Westinghouse PWR-1200 MW nuclear reactor. The proposed controller’s performance is evaluated in three scenarios:sudden control rod withdrawal, sudden change in coolant inlet temperature, and linear load tracking. The influences of model parametric uncertainty are also investigated. Simulation results revealed that the performance of the control algorithm applied using variable coolant flow rate was superior to the one usingcontrol rod positioning.

Analysis of the thermal-mechanical behavior of SFR fuel pins during fast unprotected transient overpower accidents using the GERMINAL fuel performance code

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

A practical subcritical rod worth measurement technique based on the improved neutron source multiplication method

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subcritical states. Moreover, the noise reduction technique has been adopted to smooth the large fluctuation existing in the original signals. In order to verify the engineering performance of the proposed measurement technique, the measured source-range detector count rates during the rod withdrawal process before reaching critical state in a CNP1000 reactor have been employed. It demonstrated that almost all estimated results of control rod worth satisfy the engineering acceptance criteria, except one control rod with the relative difference over 10 %, which indicates the capability of the proposed method in estimating control rod worth.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

Fuel performance evaluation of two high burnup PWR core designs during normal operation, control rod withdrawal, and control rod ejection scenarios

There is interest among utilities to extend the current, 18-month operating cycle to 24 months. Economically, this extension would require greater than 5 % enrichment and peak rod average discharge burnup levels above 62 GWd/MTU. A notable challenge of increasing enrichment is the resulting additional excess reactivity encountered during the early stages of fuel life. To accommodate, burnable absorbers beyond soluble boron are introduced into the fuel system. In high burnup fuels, the possibilities of cladding lift-off and fuel melting increase due, in part, to increased rod internal pressures and limited fuel thermal conductivity, respectively. This work collaboratively employs PARCS, RELAP5-3D, and BISON to compare the fuel performance of two high burnup fuel candidates with higher than 5 % enrichment. The fuel performance parameters were compared to current NRC guidance. The results demonstrate an annular fuel design with homogenously blended gadolinium as a burnable absorber operates with greater safety margins during normal operation, allowing for additional operational flexibility. During normal operation, the core design utilizing Integral Fuel Burnable Absorber pins contained fuel pins which reached plenum pressures above 15.5 MPa by the end of the first fuel cycle and fuel pins experienced cladding hoop strains above 1 %. In the Gd core design, only two observed pins experienced plenum pressures above 15.5 MPa and no pins exceeded 1 % cladding hoop strain. During the control rod withdrawal scenario, plenum pressures for pins in both designs marginally exceeded system pressure, however neither experienced excessive hoop strain. The Gd core design experienced a maximum fuel temperature of 2418 K, which is significantly higher than the Integral Fuel Burnable Absorber design at 2157 K, but still within regulatory guidance. We predicted that the fuel in both could return to service after the CRW event. We also predicted that cladding would not fail during the Control Rod Ejection in either core design. Generally, the Integral Fuel Burnable Absorber core design performed with greater safety margin with regards to temperature during normal operation and the transient events. However, the Gd core design performed with greater safety margin regarding plenum pressure and hoop strain limits during normal operation and both transient events.

TRISO Burnup-Dependent Failure Analysis of a HTGR Design-Basis Accident Using BISON

This work assesses the failure behavior of the silicon carbide (SiC) layer in TRIstructural ISOtropic (TRISO) fuel particles with BISON during both steady-state and transient conditions for the MHTGR-350 design by General Atomics. A one-dimensional BISON model for a uranium oxycarbide–bearing TRISO is developed to simulate a power level of 50 mW/particles up to 12.4% fissions per initial metal atom (FIMA). Stress predictions for the materials of interest are presented as a function of burnup, along with the SiC failure probability computed using Weibull statistics under the assumption of realistic SiC quality. A design-basis accident (DBA) that involves control rod withdrawal is then simulated in BISON at various burnup levels. Boundary conditions during the DBA are obtained from RELAP5-3D. The predicted SiC failure probability is 3 × 10−14%. This value does not increase during the transient as a function of burnup since irradiation induces a compressive hoop stress state that reaches a maximum absolute value of around 300 MPa. This compressive stress compensates the tensile loading conditions introduced by the transient. The tensile components appear amplified at higher burnups since they increase from 100 to 140 MPa going from 6% FIMA to 12.4% FIMA. Nonetheless, burnup provides a negligible impact on the overall failure probability predictions for SiC, as transient conditions do not translate to a stress increase toward tensile values at any of the considered burnups. Additionally, a two-dimensional (2-D) exploratory BISON model is developed to determine the effects of inner pyrolytic carbon (IPyC) crack formation on SiC. IPyC failure is predicted using Weibull statistics at 105 days from the beginning of irradiation, which is set as the time for crack initiation using XFEM in the 2-D BISON model. Stress concentrations induced by the crack cause the SiC failure probability to increase with respect to the intact particle case by approximately 10 orders of magnitude. Maximum stress concentrations are found at the time of crack formation, after which stress relaxation is observed during remaining steady-state irradiations and subsequent transient simulations. Considerations are made on the results of the 2-D model being contingent on both mechanical and anisotropy properties for pyrolytic carbon. Recommendations are provided for future work involving higher dimensionality models, further investigations on uncertainties, material properties, and additional designs, such as the fluoride salt-cooled high-temperature reactor, that operate at higher burnups.

Neutronic and thermal-hydraulic calculations under steady state and transient conditions for a generic pebble bed fluoride salt cooled high-temperature reactor

Here, we develop multiphysics coupled neutronics/thermal-hydraulics models for a recently available generic Fluoride-salt-cooled High-temperature Reactor (FHR) at equilibrium core burnup conditions with reactivity control mechanisms. We utilized the pebble bed reactor transient tool KP-AGREE with multigroup homogenized cross-sections generated by the continuous energy Monte Carlo neutron transport code Serpent to simulate transient events for an FHR. Comparing global reactor parameters and safety characteristics between Serpent and published results to standalone neutronic calculations in KP-AGREE verifies our initial conditions for our transient models. This work explores three separate transients: a loss of flow accident, an overcooling accident, and a total control rod withdrawal accident. The boundary conditions of these scenarios come from either transient benchmark definitions of a gas-cooled pebble bed reactor or transient analysis of a similar FHR design. For the loss of flow accident, our predicted peak fuel and coolant outlet temperatures show a similar response to published values, albeit slightly higher, with differences between values attributed to our modeling assumptions. The overcooling transient simulation predicts a 52 K temperature increase in the maximum fuel kernel temperature and an approximately 32.6% increase in total core power due to the negative coolant temperature feedback of the FHR design. Our control rod removal transients predicted a 235 K increase in maximum fuel kernel temperature and a 107.5% increase in total core power. For all transient cases, the fuel temperature remained significantly lower than the temperature limit, but coolant outlet temperatures for the reactivity-initiated accident exceeded external core material temperature limits.

TRISO burnup-dependent failure analysis in FHRs using BISON

In this study, we perform a study on TRIstructural ISOtropic (TRISO) Silicon Carbide (SiC) layer failure using the BISON fuel performance code. SiC thermomechanical behavior is analyzed during both steady-state and accident scenarios pertaining to a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). A monodimensional single particle model is adapted from a previous analysis and run at 140 mW/particle for 500 days. The maximum burnup reached is 18.49% FIMA. A total of three accident scenarios are simulated at both fresh fuel conditions and following multiple irradiations at 2% FIMA increments up to 18.49% FIMA. The instances include an overcooling transient, a Control Rod Withdrawal (CRW) and a Loss of Forced Cooling Accident (LOFA). SiC hoop stress and failure probability predictions are evaluated for all simulated cases. A consistent tangential compressive stress state is found as a function of burnup for SiC. The maximum compressive hoop stress reaches approximately -400 MPa. The maximum SiC failure probability as a function of burnup is 2⋅10−13% at steady-state and no increases are predicted during transient simulations. It is concluded that the selected accident scenarios, under our modeling assumptions, are not predicted to pose a challenge to SiC integrity due to their relatively slow progression and low temperature increments, for both fresh fuel and at burnup values up to 18.49% FIMA. A 2D model is used, along with XFEM, to simulate a crack in the Inner Pyrolytic Carbon (IPyC). Conservative IPyC failure predictions are obtained with Weibull statistics approximately 38 days after beginning of irradiation. After crack formation, hoop stress concentrations are predicted in SiC using conservative assumptions, with failure probability increasing to 3.3⋅10−4%. 2D BISON simulations of the base irradiation are then carried out to the maximum burnup and transient conditions are subsequently applied. SiC maximum tensile hoop stress is predicted at the time of crack initiation, after which crack expansion reduces the predicted stress magnitude in the material. The results presented in the work are interpreted under the selected modeling assumptions and conclusions are based on a comparison with a parallel analysis on a High-Temperature Gas-Cooled Reactor design to evaluate the effects of power density and burnup on SiC performance.

Advances for the time-dependent Monte Carlo neutron transport analysis in McCARD

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN-201K.

Statistical learning model for fuel-cladding heat exchange coefficient evaluation in SFR

Multiphysics modeling has become a central point in nuclear reactor simulation as it takes into account interactions between physical fields involved. Nuclear reactors are highly multi-physics systems as neutronics, thermal transfer, mechanics and hydraulics are interacting to produce and maintain the power. In this system, numerous issues come from material behavior. In the case of Control Rod Withdrawal (CRW) accident the main issues come with fuel behavior as the major risk to prevent is the fuel melting and spread in the core, leading to its partial or complete meltdown. In these situations, specialized codes are often used to predict properties and the state of the fuel by aggregating isolated models, each one corresponding to a single phenomenon. This type of approach is very efficient for understanding global sequence of the accident and predicting numerous physical variables of the problem. However it is often very expensive in calculation time. In this paper we focus on developing a fast tool based on machine learning models in order to speed up the calculation of specific variable of interest. We propose here a brief physical context of the study, a description of the method used to chose and to train models and finally an evaluation of models. Using this tool in a multiphysics scheme will add negligible penalty on calculation time. The model is focused on modeling fuel cladding heat exchange coefficient hgap based on data extracted from GERMINAL-V2 fuel performance code developed at CEA for SFR. Considering the strong non-linearity of the variation of the hgap, it appears that Random Forest models and AdaBoost present the smallest deviation, lower than 1% and the faster response, less than 10μs. However, the sensitivity of the hgap for highest burnups does not allow to get a metamodel with a deviation smaller than 20% for the application case.

A comprehensive Thermo-hydraulic neutronic and safety analysis of a 100MWth pebble bed reactor core

The HTMR100 is a 100MWth High Temperature Gas-cooled Reactor (HTGR), which is helium-cooled and graphite moderated. This reactor is of the Pebble Bed type featuring a Once-Through-Then-Out (OTTO) fuelling scheme, this cycle was chosen for the simplicity of the fuel handling system.The HTMR100 Pebble Bed Reactor (PBR) has a core volume of 27.73 m3 (core diameter 2.6 m, core height ∼ 5.2 m) which results in an average core power density of 3.6 MW/m3. This reactor utilizes uranium dioxide fuel (UO2) at a U-235 enrichment of 10 wt% and 10 g of heavy metal (total uranium) per Fuel Sphere (FS). Although there are various types of fuel options that can be utilized in the HTMR100 reactor, ranging from LEU to mixtures of Th/LEU, Th/HEU and Th/Pu, this study will only focus on LEU in the form of UO2 fuel.This study investigates the neutronic and thermal–hydraulic modelling of the 100MWth HTMR100 at equilibrium (100% power with the control rods situated at the nominal position for a core having an equilibrium fuel burnup profile) and at transient/accident conditions. The purpose of the investigation is to determine if the maximum fuel temperature remains below the set limit of 1600 °C during a design-base accident event that is defined as a Loss of Coolant Accident (LOCA).The equilibrium core was simulated using the Very Superior Old Programs (VSOP99) suite of codes to obtain burnup results. The transient behaviour of the core was modelled by a code called Multi-group Time Dependent Neutronics and Temperatures (MGT). MGT assesses the dynamic transient results of the accident scenarios such as the Depressurized Loss of Forced Cooling (DLOFC), Pressurized Loss of Forced Cooling (PLOFC), Control Rod Withdrawal (CRW) and a Control Rod Ejection (CRE) as well as normal operating transients of which a Load Following (LF) operation is a typical example.The HTMR100 reactor utilizing the enrichment and heavy metal loading as specified does indeed produce the targeted 80 000 MWD/THM burnup for the OTTO fuel cycle. The study also proves that the VSOP99 and the MGT codes do in fact yield similar results for the HTMR100 reactor with regards to fuel centreline temperatures, outer sphere surface temperatures as well as moderator temperatures for the postulated accident scenarios that were analysed. The results also indicate that the design-base transient safety analysis simulations prove that the fuel temperatures remain below the set value of 1600 °C for the oxide-based fuel. A temperature-volume analysis of a DLOFC design-base event shows that only ∼ 2 % of the fuel reaches the higher temperatures of 1500 °C to 1600 °C in a DLOFC event.The beyond-design base events with a very low probability of occurrence do however exceed the set value of 1600 °C but the fuel only exceeds this set limit for short periods of time while the transient occurs. The probability that this will lead to fuel damage should be very small since the time at these high temperatures is short and the volume fraction of the core exposed to these high temperatures is small.

Sensitivity analysis applied to SiC failure probability in TRISO modeled with BISON

A sensitivity analysis on the failure probability of the Silicon Carbide (SiC) layer in tristructural isotropic (TRISO) nuclear fuel during transient conditions predicted by the BISON fuel performance code is performed. The principal goal of the analysis is to understand the most important parameters dictating SiC failure behavior in BISON during Reactivity Initiated Accidents (RIAs). SiC brittle fracture probability is modeled using Weibull statistics. A total of seven inputs related to SiC failure has been selected for the analysis, including the Weibull statistics parameters, elastic moduli for SiC and Pyrolitic Carbon (PyC) and SiC stress-free temperature. A 1D TRISO BISON model has been established for various reactivity insertions performed at the Nuclear Safety Research Reactor (NSRR). The principal advantage associated with the 1D TRISO model developed in this work is its computational efficiency. The Sobol variance decomposition method is used, and the sensitivity indices are presented for eight different values of the energy deposition. The results show that the two most important parameters impacting the predicted SiC failure probability are the Weibull modulus and the characteristic stress, and a co-variance amongst these parameters is obtained for low reactivity insertions. An additional new finding of this work is that the relative importance of Weibull parameters depends on the energy deposition, and thus reactivity, regime. For low energy depositions, two parameters are of influence on SiC failure probability, while for high energy depositions only one parameter impacts failure probability results. Moreover, optimization of the Weibull modulus and characteristic stress is performed by minimizing the RMSE between BISON failure probability predictions and experimental failure fractions for each energy deposition. This work also demonstrates the validity of the NSRR tests BISON simulations and of the respective sensitivity analysis results as conservative, yet indicative for slower transients characterized by lower deposited energies. Such verification is achieved through the partial extension of the analysis to a group Control Rod Withdrawal reproduced from a previous study. Another novel result of this analysis is that no single set of Weibull parameters can reproduce all reactivity insertion experimental failure results, which is related to the intrinsic nature of SiC failure properties, and a new range for the parameters is proposed to produce a failure probability envelope that encompasses the experimental fractions. Additionaly, this work proposes a new approach for future failure analysis with BISON, consisting in the use of Weibull parameters ranges, rather than fixed sets, along with failure envelopes generation.

TRISO SiC Failure Probability for Reactivity Initiated Accidents in High-Temperature Gas-Cooled Reactors

This work analyzes the failure process of the silicon carbide (SiC) layer in tristructural isotropic (TRISO) during reactivity-initiated accident scenarios for a high-temperature gas-cooled reactor (HTGR) with BISON. Two cases are considered—a group control rod withdrawal (CRW) and a control rod ejection (CRE)—reproduced from a previous study. Failure probability is modeled using Weibull statistics, and worst-case scenario Weibull parameters are adopted to simulate the envelopes in BISON with a one-dimensional TRISO model. CRW scenario results are characterized by higher values of maximum energy deposition and final temperature and volumetric strain with respect to the CRE ones, but the latter have remarkably higher SiC failure probability, mainly due to the offset in strain rates between the two cases. This work also confirms the validity and conservatism of the performance envelopes produced in a previous work by replicating the envelope formulation using RELAP5-3D and RAVEN with a different sampling technique and obtaining consistent results. A sensitivity analysis using the Sobol variance decomposition method on SiC failure probability is then performed involving a set of inputs on both CRW and CRE. The two most important parameters are Weibull modulus and characteristic stress, and their relative importance depends on the specific case. The proposed interpretation of the results is that both energy deposition and strain rate influence the relative degree of importance of the failure parameters. Computation of 95% confidence intervals around worst-case scenario SiC failure probability values is also carried out for four different sets of Weibull parameters. A new criterion for SiC TRISO quality classification built upon safety-based ranges of Weibull parameters is proposed to be integrated in future Fuel-Production Quality Assurance Plans.

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

The system code ATHLET (Analysis of THermal-hydraulics of Leaks and Transients) is being developed by the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) mbH in Garching, Germany. In the paper, an overview of activities performed at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, in application of system code ATHLET in transient analyses for NPP Krško (NEK) is presented. Newly, the NEK input deck for the released ATHLET version (Mod 2.2 Cycle A) has been developed. For that purpose, the NEK data base that has been developed and maintained at FER for the last two decades primarily for development of standard input deck for RELAP5 code was used. The ATHLET model has been validated by analyzing the Rod Withdrawal At Power (RWAP) accident at nominal power. The results for steady state calculation as well as RWAP transient were assessed against the analysis performed by RELAP5/mod 3.3 code. In both ATHLET and RELAP5 calculation, the RWAP accident was simulated by constant reactivity insertion rate equal to 2.4 pcm/sec. For ATHLET analysis, two fluid dynamic options were tested for the primary side: a) base case analysis with 5 conservation equations and mixture level model and b) two-fluid model with separate conservation equations for liquid and vapour phases for all the volumes except for the pressurizer where 5 equations+mixture level model was retained. The Steam Generators (SGs) were built using basic ATHLET elements together with the dedicated separator model. For RELAP5/mod 3.3 analysis, a standard option with thermal and mechanical non-equilibrium (6 equations) was used. The results of the steady state calculation for the ATHLET model have shown a very good agreement with RELAP5 calculation. In the transient analysis very small differences for the main physical parameters between ATHLET and RELAP5 as well as between the two ATHLET models were obtained.

Pin-by-Pin Coupled Transient Monte Carlo Analysis Using the iMC Code

In this article, we present a coupled multi-physics Monte Carlo reactor transient analysis framework implemented in the KAIST Monte Carlo iMC code. In the multi-physics framework, the time-dependent neutron transport calculation and the transient heat transfer analysis are done based on the predictor–corrector quasi-static Monte Carlo method and the three-dimensional finite element method, respectively. Using this high-fidelity analysis framework, we demonstrated the negative temperature feedback effect in two pressurized water reactor (PWR) transient scenarios. First, a 3-D burnable absorber-loaded fuel assembly was considered with all reflective boundary conditions. In this simple problem, a positive reactivity-induced transient was analyzed to characterize the reactor responses in view of the pin-wise power and temperature distribution. Second, the iMC multi-physics analysis is applied to a control rod withdrawal transient in the TMI-1 mini core problem, and detailed time-dependent results were provided and compared with the Serpent/SUBCHANFLOW analysis. In both cases, independent MC runs were performed to quantify the uncertainty of the multi-physics MC transient analysis.

Optimized fractional-order PID controller based on nonlinear point kinetic model for VVER-1000 reactor

Abstract Nuclear reactor dynamics are nonlinear and time-varying, so the power level control is a challenging problem in nuclear power plants (NPPs) to ensure both its operation stability and efficiency. An important measure to improve the safety of the reactor core of NPP is the implementation of robust control for the core by adjusting the inserted reactivity of the control rods. Thus in the present paper, fractional-order PID (FOPID) controller is developed as it is well known for its simplicity and robustness against disturbances. A Genetic Algorithm (GA) is used to determine FOPID controller parameters to achieve the desired power level for the generation III+ reactor VVER-1000. Implementing the GA, a suitable objective function is proposed to search for the optimal FOPID parameters. The nonlinear model of the VVER-1000 nuclear reactor is presented based on the point kinetic equations with six delayed neutron groups and temperature feedback from lumped fuel and coolant temperatures. Two cases for the VVER-1000 reactor are investigated; the changes in the power loads and the control rod withdrawal that leads to reactivity disturbance. Moreover, the uncertainties that result from model parameters perturbation are added to examine the controller robustness. The simulation results show that the proposed optimized FOPID controller can track the desired power level of the VVER-1000 reactor and robustly cope with any load changes, disturbances, or any parameters uncertainties. Also, it proves the superiority of the proposed optimized FOPID controller over other PID controllers in ensuring the safe and effective operation of the VVER-1000 reactor.

Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

Additional Stylized 3-D Benchmark Problem Set Based on the Pin-Fueled SmAHTR

Previous work presented a set of stylized three-dimensional benchmark problems based on the Oak Ridge National Laboratory (ORNL) preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, with prismatic assemblies fueled by tri-isotropic (TRISO) particles. That previous work created a detailed description of the benchmark problems by closing several outstanding design gaps from the ORNL preconceptual design report, notably by addressing the lack of active control mechanisms, for which control rod “bundles” were implemented. In this paper, the creation of two additional stylized benchmark problem sets based on that past work is detailed, offering two new control rod configurations. The fluoride salt, small size, and highly heterogeneous TRISO-fueled pins make these additional benchmark problem sets useful numerical validation references in benchmarking neutronics tools against continuous-energy stochastic Monte Carlo results. Detailed reference results, including the eigenvalue (keff) and 1/11th assembly-averaged relative fission density distributions, are provided for both control rod configurations in full-core cases with all control rods withdrawn and all control rods fully inserted. A near-critical core benchmark problem and results are provided for one configuration. The provided results are calculated using the continuous-energy Monte Carlo code MCNP.