Control Rod Banks Research Articles

On excess reactivity control of the advanced high temperature reactor (AHTR)

The Advanced High Temperature Reactor (AHTR) is a large Fluoride salt cooled High temperature Reactor (FHR) with a thermal power of 3400 MWt, utilizing hexagonal fuel elements with TRISO fuel spheres embedded in graphite fuel plates. While its cycle length is relatively short (less than one year), its specific power is high, resulting in high cycle burnup and consequently high initial excess reactivity that needs to be controlled. This is achieved by using both Eu2O3 burnable poison spheres and molybdenum-hafnium-carbon control rods. This work details studies developing and implementing these two reactivity control features over representative AHTR fuel cycles. Various Eu2O3 densities are considered to evaluate both the achieved reactivity control and the residual reactivity penalty of burnable poison at end of cycle. Control rod studies are performed to find integral and partial insertion reactivity worth for control rod banks at various radial locations. Axial power impacts are also evaluated. Focus then shifts toward control rod movement strategies. A manually generated insertion scheme leading to a critical configuration is first investigated to serve as a reference point for automated schemes to follow. Next, automated insertion and withdrawal strategies are evaluated based on their radial power peaking performance. Control rod insertions into the core locations with the highest local power are shown to be more stable than control rod withdrawals from core locations with the lowest local power. Finally, a methodology is introduced for finding a critical control rod insertion configuration. The methodology is intended to be used over a depletion sequence where the positions of the control rods are automatically adjusted to maintain criticality over cycle. In summary, this work presents a possible implementation of AHTR reactivity control and introduces automated control rod movement strategies for use in multi-physics depletion analyses.

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Deep Learning Approach to Within-Bank Fault Detection and Diagnostics of Fine Motion Control Rod Drives

Control rod motion is one of the primary means of regulating the rate of fission in a nuclear reactor core to ensure safe and stable operation. Reactor power distribution and thermal power output can be fine-tuned by adjusting the control rod position. For high-precision control of rod movements, Fine Motion Control Rod Drives (FMCRDs) are often used. The operation of FMCRDs provides a unique opportunity to implement condition monitoring related to the intermittency of motion and the use of control rod banks. This research sets out to detect three types of faults in an electrically driven FMCRD. In addition to detecting faults, this work will attempt to determine both the type of fault and the source of each fault, completing the fault detection and diagnostics (FDD) pipeline on a scarcely researched system. The three types of faults to be investigated are short-circuit faults, ball screw wear faults, and ball screw jam faults. This is a potential advancement to the within-bank FDD of this specific drive system intended for deployment in an advanced nuclear reactor plant. Using encoder-decoder structured convolutional neural networks and autoencoders, the three tested faults were confidently detected and isolated as well as reasonably diagnosed by monitoring the FMCRD servomotor torque.

Whole core analysis of BEAVRS benchmark for hot zero power condition using MVP3 with JENDL-5

ABSTRACT With the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-measurement (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross-section data from the JENDL-4.0u1 with JENDL-5 revealed that 1H, 235U, 238U, and 16O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated compared with the measured values, whereas those of with 2.78 K were improved and agreed with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.

Establishment of DeCART/MIG stochastic sampling code system and Application to UAM and BEAVRS benchmarks

In this study, a DeCART/MIG uncertainty quantification (UQ) analysis code system with a multi-correlated cross section stochastic sampling (S.S.) module was established and verified through the UAM (Uncertainty Analysis in Modeling) and the BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) benchmark calculations. For the S.S. calculations, a sample of 500 DeCART multi-group cross section sets for two major actinides, i.e., 235U and 238U, were generated by the MIG code and covariance data from the ENDF/B-VII.1 evaluated nuclear data library. In the three pin problems (i.e. TMI-1, PB2, and Koz-6) from the UAM benchmark, the uncertainties in kinf by the DeCART/MIG S.S. calculations agreed very well with the sensitivity and uncertainty (S/U) perturbation results by DeCART/MUSAD and the S/U direct subtraction (S/U-DS) results by the DeCART/MIG. From these results, it was concluded that the multi-group cross section sampling module of the MIG code works correctly and accurately. In the BEAVRS whole benchmark problems, the uncertainties in the control rod bank worth, isothermal temperature coefficient, power distribution, and critical boron concentration due to cross section uncertainties were calculated by the DeCART/MIG code system. Overall, the uncertainties in these design parameters were less than the general design review criteria of a typical pressurized water reactor start-up case. This newly-developed DeCART/MIG UQ analysis code system by the S.S. method can be widely utilized as uncertainty analysis and margin estimation tools for developing and designing new advanced nuclear reactors.

High-Performance and High-Fidelity GPU-Based Monte Carlo Solutions to the BEAVRS Benchmark

The BEAVRS benchmark is solved by PRAGMA, the graphics processing unit (GPU)–based continuous-energy Monte Carlo code. The solutions consist of the detailed simulation results for the two cycles that involve the reactivity and pin power distribution information for the zero-power physics tests and depletion. Primary results at hot zero power, such as the critical boron concentration at various rodded conditions, control rod bank worth, isothermal temperature coefficients, and assemblywise detector signal, are compared with the measured data. Core-follow calculations are performed with varied power, and the resulting boron letdown curves are compared with the measured one. Hot full-power depletion is also performed and the resulting pinwise power distributions of cycle 1 are compared with the nTRACER results. The comparison with the measured data and also with the nTRACER results demonstrates the high solution fidelity of PRAGMA. In all the calculations, PRAGMA uses a tremendously large number of histories, ranging from up to hundreds of millions per cycle, that are used to fully exploit the massive parallel computing capacity of GPUs. The execution time of the entire core-follow calculation with about 30 burnup steps takes less than 16 h on a single rack of computing nodes mounted with 24 gaming GPUs, which represents considerably high Monte Carlo core calculation performance.

MEASURED AND CALCULATED INTEGRAL REACTIVITY OF CONTROL RODS IN RSG-GAS FIRST CORE

The control rod worth is one of the important parameters for the operation of a nuclear reactor. Proper measurement and calculation of the control rod worth are essential for the safe reactor operation under normal and transient conditions that are initiated by a postulated event such as a stuck rod, control rods ejection, etc. This paper presents calculation results of integral reactivity of the RSG-GAS research reactor first core and its comparison with the experimental data. Calculations were performed using the continuous energy transport code Serpent 2 with ENDF/B-VIII.0 nuclear data. Integral reactivity measurement was done by compensating method with control rod bank, regulating rod, and reactivity meter. Calculations are carried out for each method used in control rod measurement data with an aim to validate calculated results to experimental data. Compared with the measured experiment data, there are no significant differences in calculation results of integral reactivity. The maximum difference of the control rod's total reactivity is 1.26% compared to the measurement carried out by compensating method with regulating rod.

Study of rod ejection accident at hot zero power condition in a PWR using RELAP5

The core reactivity and power distribution within the core of a pressurized water reactor (PWR) are controlled for safe and efficient operation of the reactor. Several postulated events, however, can alter the core reactivity and power distribution and disrupt reactor operation. Such those events are the reactivity accidents. Among the different expected reactivity accidents, the control rod ejection accident (REA) is of a particular concern. This accident can occur in PWRs by mechanical failure of the control rod drive mechanism or due to inadvertent operator action. There are two cases of REA; REA at hot zero power condition, and REA at power condition. This accident results in reactivity addition, and based on the rate of reactivity addition, the reactor power may increases to levels large enough to endanger the fuel assemblies faster than the automatic protective actions.In the current work, RELAP5 code is used to study the consequences of REA at Hot Zero Power Condition in a standard four loop Westinghouse PWR reactor. A reactivity insertion of up to 75 pcm/sec is assumed in the analysis due to the rod movement. This value is consistent with the withdrawal of two control rod banks. A RELAP5 input file is prepared to simulate the reactor core and primary system. In this file, the core is simulated by two channels, a hot channel represents a one fuel assembly and an average channel represents the rest of the core. RELAP5 reactor kinetics model is used to determine the power generation with time including the core feedback effects; i.e., Doppler reactivity and moderator reactivity. The reactor is set to trip when it reaches 35% of full power, the low setting reactor trip assumed to be initiated by power range high neutron flux. The initial power level was assumed to be below the power level expected for any shutdown condition (10−9 of nominal power).Results of the study including the peak power, maximum fuel pellet temperature and fuel pellet enthalpy are presented. The effects of the rod ejection rate on REA consequences are also studied.

Control rod shadowing effect in PWR core utilizing Urania-Gadolinia fuel

Safety analysis of nuclear power plants requires the assessment of all the systems important to safety. Control rods are one of the important safety systems for nuclear reactors. The control rod worth is affected by location of control rod banks, enrichment of uranium fuel in the assembly where the bank is inserted and the enrichment of gadolinium in the assembly. Also, the Interactions between control rods, phenomena known as shadowing effect, influence the worth of control rods. In this paper shadowing effect and the factors affecting the control rod worth are investigated for four types of control rod banks of an EPR Reactor. Parameters investigated are control rod worth, degree of interaction between rods and amplification factor. Moreover, the effect of interactions between control rod banks on the fluctuation of flux distribution in the core is studied. The calculations are evaluated at BOC and EOC using MCNP6 code. The results showed that the location of control rod bank in the core has the highest effect on the worth of a certain bank, second is the number of control rod bank groups in the core, followed by the fuel enrichment. The gadolinium burnable absorber has a slight effect on the control rod worth. The results obtained represent an important contribution that would support safety analysis of the EPR reactor design.

Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

The BEAVRS benchmark provides detailed design data and in core flux measurements for a standard PWR. In this work the BEAVRS benchmark is simulated at hot zero power using the Monte Carlo code MCNP6. The effective multiplication factor is estimated at various control banks insertions and boron concentrations. Calculations also include control rod banks worth, and isothermal temperature coefficient. Axially integrated Thermal flux in 58 assemblies, resembling detector positions in the core, are also evaluated and compared to the actual data provided by the benchmark. Axial thermal flux calculated for selected assemblies was compared to the results produced by detector signals located at 61 axial positions. Radial power distribution in the whole core is also evaluated, and compared to a previous study. The accuracy of thermal flux and radial power calculations were evaluated using two methods; the absolute relative difference and the root mean square deviation. The model was capable of simulating the benchmark with good degree of accuracy.

ANALYSIS OF THE BEAVRS BENCHMARK USING CASMO5/SIMULATE5 WITH JENDL-4.0 AND ENDF/B-VII.1

The BEAVRS benchmark was analyzed using the CASMO5/SIMULATE5 in order to compare the measurement data and the calculation results based on the JENDL-4.0 and ENDF/B-VII.1 and investigate the difference between those calculation results. For the hot zero power (HZP) physics test, the calculation results showed good agreement with the measurement data for both of cycles 1 and 2. For cycle 1, the calculation results of the isothermal temperature coefficient (ITC) differed from the measurement data by approximately 1 pcm/℉, and the same tendency has been reported in previous studies. For the cycle operation, the calculation results of the boron letdown agreed well with the measurement data. On the other hand, some calculation results of the axial detector signals had a large difference from the measurement data, which is supposedly attributed to the discrepancy of the axial offset (AO) caused by the authors’ approximation for the control rod and shutdown bank positions. In terms of the comparison between the JENDL-4.0 and ENDF/B-VII.1, although approximately 15 ppm difference of the boron letdown in the cycle operation was observed, no significant difference was seen for other core parameters, thus, the influence of the two nuclear data library was small on the present results.

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.

A comparison of three algorithms applied in thermal-hydraulics and neutronics codes coupling for lbe-cooled fast reactor

In this research, the Operator-Splitting semi-implicit (OSSI) method, fixed-point implicit (FPI) method and Approximate Block Newton (ABN) method have been applied to the coupling of thermal-hydraulics code COBRA and neutronics code SKETCH-N. In the OSSI method, there’s only one data exchange within a time step. In the FPI coupling, an iteration convergence has been added to the OSSI coupling framework to ensure the coupling results converged. And in the ABN method, a variant of Jacobian-free Newton Krylov method has been applied, which removes the Jacobian matrix construction and storage. Three temporal schemes have been validated through a PWR control rod bank ejection benchmark to show their applicability. Moreover, a LBE-cooled fast reactor control rods withdrawal accident applying two time steps has been simulated. And the results indicate that the ABN method applied in the mild transient generally outperforms the other two methods in both the convergence speed and the computational cost.

Application of deep neural networks for high-dimensional large BWR core neutronics

Compositions of large nuclear cores (e.g. boiling water reactors) are highly heterogeneous in terms of fuel composition, control rod insertions and flow regimes. For this reason, they usually lack high order of symmetry (e.g. 1/4, 1/8) making it difficult to estimate their neutronic parameters for large spaces of possible loading patterns. A detailed hyperparameter optimization technique (a combination of manual and Gaussian process search) is used to train and optimize deep neural networks for the prediction of three neutronic parameters for the Ringhals-1 BWR unit: power peaking factors (PPF), control rod bank level, and cycle length. Simulation data is generated based on half-symmetry using PARCS core simulator by shuffling a total of 196 assemblies. The results demonstrate a promising performance by the deep networks as acceptable mean absolute error values are found for the global maximum PPF (~0.2) and for the radially and axially averaged PPF (~0.05). The mean difference between targets and predictions for the control rod level is about 5% insertion depth. Lastly, cycle length labels are predicted with 82% accuracy. The results also demonstrate that 10,000 samples are adequate to capture about 80% of the high-dimensional space, with minor improvements found for larger number of samples. The promising findings of this work prove the ability of deep neural networks to resolve high dimensionality issues of large cores in the nuclear area.

BEAVRS benchmark analyses by DeCART stand-alone calculations and comparison with DeCART/MATRA multi-physics coupling calculations

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark calculations were performed by DeCART stand-alone and DeCART/MATRA multi-physics coupled code system to verify their accuracy. The solutions of DeCART stand-alone calculations for the control rod bank worth, detector signal, isothermal temperature coefficient, and critical boron concentration agreed very well with the measurements. The root-mean-square errors of the boron letdown curves for two-cycles were less than about 20 ppm, while the individual and total control rod bank worth agreed well within 7.3% and 2.4%, respectively. For the BEAVRS benchmark calculations at the beginning of burnup, the difference between DeCART simplified thermal-hydraulic stand-alone and DeCART/MATRA coupled calculations were not significantly large. Therefore, it is concluded that both the DeCART stand-alone code and the DeCART/MATRA multi-physics coupled code system have the capabilities to generate high fidelity transport solutions at core follow calculations.

A correction method for RAPID fission matrix calculations with control rod movement

The RAPID radiation transport code, based on a fission matrix combination method, was initially designed for spent fuel systems but has been applied on reactor core neutronics analysis. To improve the accuracy of the RAPID fission-matrix approach, a ratio correction method has been applied to the reflector region and for heterogeneous adjacent assembly types. The RAPID code has been tested on a modified BEAVRS benchmark at hot zero power condition without control rods, and reports accurate results (∼0.52% 2-D, pin-wise RMS error with ∼25 pcm keff difference compared to a Serpent 2 reference). In this paper, we apply the assembly correction to rodded assemblies on different control rod bank worth calculations, as well as separate ratios on different slices in order to solve for the partially inserted control rod problems. It is shown that RAPID is able to estimate the control rod worth for different banks (a maximum 37 pcm error and most within 20 pcm). RAPID can also perform calculations on a partially inserted control rod case, reporting a maximum control rod worth error of 17 pcm across some insertion levels of bank D. In addition, another type of correction ratio method is proposed, namely the “localized” ratio. It makes use of sets of four-assembly fission matrices from criticality calculations and allows the choice of any fission source in ratio generation. It greatly reduces the errors located near large source gradient locations, which was observed with the “multiplied” correction ratio since it always assumes a uniform source.

Uncertainty propagation analysis for control rod worth of PWR based on the statistical sampling method

Based on the idea of uncertainty propagation, global relative covariance matrices of few-group cross sections representing the uncertainties propagated from the basic nuclear data and manufacturing tolerance were built by using statistical method. Then the twice critical calculation coupled with efficient sampling (TCCES) method based on the innovative efficient sampling method of Latin hypercube sampling combined with singular value decomposition conversion (LHS-SVDC) was chosen to quantify the CRW uncertainty. A simplified Mini-core model in a 3 × 3 checker board pattern based on the CASL benchmark and the steady-state BEAVRS benchmark were chosen for the CRW uncertainty analysis. For the purpose of verifying the modeling and simulation, lattice calculations, keff calculation and control rod worth calculation for different types of control rod banks were all performed and compared with the published results obtained by other MC codes and deterministic method codes. Finally, the uncertainties of the differential and integral CRW for different types of control rod banks were quantified and the contributions of nuclear data and manufacturing tolerance were also identified.

A decoupled Mechanical Shim core control strategy for a pressurized water reactor using feedforward compensation and a multimodel approach

The advanced Mechanical Shim (MSHIM) core control strategy employs two separate and independent control rod banks, namely the MSHIM control banks (M-banks) and axial offset (AO) control bank (AO-bank), for automatic reactivity/temperature and axial power distribution control respectively. The M-banks and AO-bank are controlled by two closed-loop controllers, independently, called the coolant average temperature (Tavg) controller and the AO controller. Since the motion of the M-banks and AO-bank can both affect the Tavg and AO, an interlocking design between the Tavg controller and the AO controller is adopted in the MSHIM control strategy to avoid the interference between the two controllers during power maneuvers. This design can enhance the stability of the MSHIM control system by avoiding the simultaneous movement of the M-banks and AO-bank and keeping the priority of the M-bank movement. However, the AO control performance is degraded at the same time. In the present study, a feedforward compensation decoupling method and a multimodel approach have been used to eliminate the coupling effect between the two controllers in the MSHIM control system during a wide range of power maneuvers. A multiple feedforward compensation system has been designed with integration of the feedforward compensators for the Tavg and AO controllers at five equilibrium conditions using the multimodel approach. By implementing it in the MSHIM control system, the interlocking between the M-banks and the AO-bank can be released to realize the independent and decoupling control between the Tavg and AO. The effectiveness of the decoupled MSHIM control strategy has been verified by comparing its control performance with that of the original MSHIM control strategy during typical load change transients of the AP1000 reactor. The obtained results show that superior and decoupling control of the Tavg and AO can be achieved using the proposed decoupled MSHIM control strategy.

PWR advanced fuel high burnup-dependent xenon oscillation controllability assessment

The controllability of xenon oscillations in a Pressurized Water Reactor (PWR) fuel assembly was evaluated using the MCNP and ORIGEN-2 codes. Burnup-dependent symmetrical cosine depleted nuclide profiles were generated for advanced high burnup uranium oxide (UO2) and mixed oxide (MOX) fuels. A Monte Carlo burnup coupling BASH script (MCOS) was used to calculate burnup-dependent neutron Fission Tally and 135Xe axial relative oscillation profiles. Burnup-dependent xenon oscillation controllability was then assessed for the advanced high-burnup UO2 and MOX fuels using MCOS-calculated xenon reactivity worth swing bands (ΔK) with respect to the mean. The advanced fuels xenon oscillation bands ΔK are bounded by the typical PWR uranium oxide fuel with 235U enrichment of 3.0 wt%. It has demonstrated that the typical PWR full/partial length control rod banks system’s reactivity can adequately control the high burnup advanced fuels xenon oscillation in full power operation.

Validation of SuperMC with BEAVRS benchmark at hot zero power condition

SuperMC is a general, intelligent, accurate and precise simulation software system for the nuclear design and safety evaluation of nuclear systems. In this study, the latest version of SuperMC has been validated by Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), which is a full-core commercial Westinghouse PWR core, including two cycles of measured operational data. The benchmark is widely used in the validation of Monte Carlo codes, because it is an extremely detailed, relevant and large-scale 3D PWR model based on a real reactor. Firstly, the detailed BEAVRS model is built with the CAD-based hierarchical modeling function of SuperMC. Secondly, several calculations using SuperMC at hot zero power condition of cycle 1 are performed, and the comparison with the reference results and measured data show great agreement, including neutron effective multiplication factor, control rod bank worth, temperature coefficients, U-235 fission rates and relative power distribution. This study demonstrates the capability of SuperMC on 3D full-core high fidelity reactor simulation.