Control Rod Research Articles

An Analysis of the Food and Drug Administration Manufacturer and User Facility Device Experience Database for MAGnetic Expansion Control Spinal Rods.

MAGnetic Expansion Control (MAGEC) rods can prevent repeated lengthening operations for scoliosis patients. However, there have been several Field Safety Notices issued, including a worldwide product recall due to actuator endcap separation. We aimed to review adverse events reported to the Food and Drug Administration (FDA) regarding MAGEC rods, focusing on MAGEC X. Reports submitted to the Manufacturer and User Facility Device Experience database in relation to MAGEC devices were accessed and analysed using R Statistical Software. Exclusion criteria included duplicate and literature review reports (n = 54). Free-text data were analysed using inductive content analysis. 1016 adverse events were reported to 11/30/2023. 99.0% (1006) were submitted by the manufacturer. Reports primarily arose from the UK (465, 45.8%) or US (421, 41.4%). From free-text data the most frequent adverse events were distraction mechanism failure (573), device wear (272), and actuator seal damage (180). Rod fracture (n = 48) was not significantly associated with rod diameter (≤ 5.0mm or > 5.0mm), p = 0.736. 234 reports referenced MAGEC X devices; actuator endcap separation was identified in 41.9% (99). Other events include failure of distraction (63), surface damage (31), and rod fracture (15). On 06/30/2020 MAGEC X2 received FDA approval. Twenty reports reference devices manufactured after this date, seven describe distraction mechanism failure; notably there are no reports of endcap separation. These data represent the largest series of adverse events reported for MAGEC rods, including significant new data regarding MAGEC X. As well as endcap separation, failure of distraction, surface damage, and rod fracture were reported.

Analysis of the influence of bottom flow holes in control rod guide tubes on flow field and control rod displacement

During the operation of pressurized water reactors, flow-induced vibration issues are commonly encountered, where control rod components vibrate due to the impact of coolant flow, leading to wear, thereby posing a threat to the safe operation of the reactor. This paper conducts a detailed numerical simulation of the flow field and force distribution inside a full-scale lower control rod guide tube based on the CFD method. The fluid–structure interaction analysis is performed to reveal the influence of cross-flow under various operating conditions on the vibration and displacement of control rods. The research subject consists of 24 control rods, one continuous guide section, and six control rod guide cards. The lower part of the guide tube has two flow holes on each surface, with one assumed to be a cross-flow inlet and the others as outlets. The results indicate that when considering the cross-flow effect at the flow holes, 90 % of the coolant flows out from other flow holes at the bottom of the guide tube, resulting in increased transverse velocity at the bottom and the presence of numerous vortices. The magnitude and location of the force exerted on the control rod by coolant impact are correlated. The control rod closest to the inlet of the flow hole experiences the greatest force, approximately 5.35 times the average at other positions. At a cross-flow velocity of 0.4 m/s and a low frequency of 10 Hz, the maximum displacement of the control rod is 0.3355 mm.

A Design and Safety Analysis of the “Electricity-Hydrogen-Ammonia” Energy Storage System: A Case Study of Haiyang Nuclear Power Plant

With the development of modernization, traditional fossil energy reserves are decreasing, and the power industry, as one of the main energy consumption forces, has begun to pay attention to increasing the proportion of clean energy generation. With the deepening of electrification, the peak-valley difference of residential electricity consumption increases, but photovoltaic and wind power generation have fluctuations and are manifested as reverse peak regulation. Thermal power plants as the main force of peak regulation gradually reduce the market share, making nuclear power plants bear the heavy responsibility of participating in peak regulation. The traditional method of adjusting operating power by inserting and removing control rods has great safety risks and wastes resources. Therefore, this paper proposes a new energy storage system that can keep the nuclear power plant running at full power and produce hydrogen to synthesize ammonia from excess power. A comprehensive evaluation model of energy storage based on z-score data standardization and objective parameter assignment AHP (analytic hierarchy process) analysis method was established to evaluate energy storage systems according to a multi-index system. With an AP1000 daily load tracking curve as the input model, the simulation model built by Aspen Plus V14 was used to calculate the operating conditions of the system. In order to provide a construction basis for practical engineering use, Haiyang Nuclear Power Plant in Shandong Province is taken as an example. The system layout scheme is proposed according to the local environmental conditions. The accident tree analysis method is combined with ALOHA 5.4.1.2 (Areal Locations of Hazardous Atmospheres) hazardous chemical analysis software and MARPLOT 5.1.1 geographic information technology. A qualitative and quantitative assessment of risk factors and the consequences of leakage, fire, and explosion accidents caused by hydrogen and ammonia storage processes is carried out to provide guidance for accident prevention and emergency rescue. The design of an “Electric-Hydrogen-Ammonia” energy storage system proposed in this paper provides a new idea for zero-carbon energy storage for the peak shaving of nuclear power plants and has a certain role in promoting the development of clean energy.

Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence

This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.

Detection of Axial Locations of Control Rods Using Movable In-core Detectors

Background Reactivity control in Pressurized Water Reactors (PWRs) relies on Rod Cluster Control Assemblies (RCCAs), or control rods (CRs), and boric acid diluted in primary circuit water. The control rods’ axial location inside the reactor core is typically provided by means of two redundant systems: the Rod Positioning Indicator (RPI) and the Rod Control System (RCS). Methods The current work focuses on the development of an additional method for the detection of control rod axial location inside the reactor core of a Pressurized Water Reactor (PWR) using the Advanced Flux Mapping System (AFMS) and its Movable In-core Detectors (MIDS). The paper presents the methodology employed and the results of the first validation campaigns. Results Two validation campaigns were carried out, targeting respectively equilibrium and non-equilibrium xenon axial distribution. The routine proved successful in identifying control rods’ axial insertion in different operating conditions including Flux Maps (FMs) acquired at 2%, 50% and 75% power. Conclusions The routine developed at Beznau Nuclear Power Plant will serve as an additional core monitoring tool to be employed in case of suspicious or faulty indication coming from the main control rod location methods. The developed method proved able to locate control rods’ axial insertion with few steps error, equivalent to maximum 1.6% of total fuel height (i.e. 5 cm as maximum error). There is room for potential improvements with regards to the uncertainty model currently implemented.

Remaining useful life prediction of nuclear reactor control rod drive mechanism based on dynamic temporal convolutional network

The control rod drive mechanism (CRDM) is a critical equipment of the nuclear reactor, and the prediction of its remaining useful life (RUL) is important for the efficient maintenance and ensuring the safe, reliable operation of nuclear power plants. In this paper, a novel framework for the RUL prediction of CRDM is proposed, which is a dynamic temporal convolution network (DTCN) based on dynamic activation function and attention mechanism. Firstly, the temporal convolution network (TCN) is used as the backbone of the prediction model, to extract the temporal dependence of the input data. Next, the dynamic activation function DReLU is integrated into the TCN, which can dynamically activate features and capture variable degradation information. Then, introducing the attention mechanism improves the influence of important high-level features extracted by the network on RUL prediction, thereby improving the efficiency of feature extraction in the network. Finally, the DTCN outputs the predicted RUL by performing non-linear mapping on the extracted features. The CRDM accelerated life test platform is established and a series of experiments are conducted using the collected CRDM full-life vibration dataset. The results demonstrated the performance and advantages of the proposed method.

Mechanism model for effect of eccentricity on capacitive control rod position indicators

The continuous identification technology for position signals based on the capacitive method has a good application potential in the field of the control rods position measurement in nuclear reactors. The main source for error of the Capacitive Control Rod Position Indicators (CCRPIs) comes from the effect of eccentricity. The conformal mapping technique was used in this paper, and the geometry of the CCRPIs could be transformed. A model for revealing the mechanism of the effect of eccentricity was established, and it was tested and verified by the method of numerical calculation. The characteristics of the distribution of the sensitive field was clarified by the model. The results indicate that the maximum error of the model is smaller than 13 %, and the eccentric effect of the indicator can be described by the model.

Performance of a hydraulic buffer for PWR fuel assemblies: Mathematical modeling, numerical solutions, and experimental comparison

In a Pressurized Water Reactor (PWR), the hydraulic buffer serves as an essential component, significantly mitigating the impact force between the control rod drive mechanism and the fuel assembly in scenarios of emergency shutdown. This paper provides a complete analysis of the dynamical performance of a new type of hydraulic buffer, including its mathematical modeling, numerical solution scheme, and experimental comparison. During the fluid modeling, five typical cases are first classified in terms of both flow directions and coefficients based on the relative positions of the sleeve and the piston. A new flow iterative calculation format encompassing flow directions that can efficiently solve the flow coefficients is proposed during the fluid modeling. A one-way coupling scheme is used for the fluid–structure dynamical solutions. An experimental comparative study is conducted using the standard and reference (SR) case, and the present method shows good agreement with the experiment. In the present analysis: (1) The dynamic characteristics of the buffer are fully demonstrated using time history curves and phase diagrams, the physical parameters of fluid and structural motions, mainly including fluid pressure inside the chamber, dynamic relative displacement and velocity of the piston and sleeve, the rebound disparagement, and system kinetic energy; (2) Typical features of double peaks of the impact force, related to the collision between the piston, the impacted object, and the sleeve, have been captured and well simulated; (3) The variation of the two peaks of the impact forces with parameters and the transformation laws of the maximum impact force between these two peaks have been fully revealed.

Uncertainty and sensitivity analysis of the fission product behaviour in the Phébus FPT1 test with the system code AC2

The Phèbus FP (Fission Product) research program, conducted by IRSN and CEA, aimed to improve the understanding of the phenomena governing core melt down, fission product release and transport, as well as the fission product behaviour in the containment. The test facility represents a 900 MWe PWR at a 1/5000 scale. The second test, FPT1, was carried out in 1996 and utilised an Ag-In-Cd alloy control rod. There is a growing interest in Best Estimate Plus Uncertainty analysis both in the research, as well as in the licensing. It has a long-standing history in case of thermal–hydraulic system codes focusing on design basis accidents, but its application is limited on the severe accident domain. In recent times however an increasing attention has been paid to evaluating severe accident codes’ uncertainties. This study presents an uncertainty and sensitivity analysis (UaSA) of the FPT1 test, focusing on fission product behavior modeling within the severe accident code system AC2. The detailed application of the coupled codes ATHLET-CD and COCOSYS in this context is novelty. Detailed description of the methodological and practical approach is provided. The uncertain input parameters chosen and quantified are directly related to the phenomena describing the fission product transport and retention phenomena both in the primary circuit, as well as in the containment, while thermal–hydraulic conditions within the primary circuit were kept as much as possible constant. Finally, the study has been complemented by a sensitivity analysis deriving the Spearman's rank correlation coefficient for each uncertain input parameter.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.

Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark

The direct whole-core calculations can provide accurate results and insights to the physics phenomena of the reactor. It can also capture the local effects of temperature and density fields on fuel depletion. However, the computational cost of the direct whole-core calculations is expensive. To compromise between computational cost and accuracy, the DDC-3D code system has been developed to perform neutronics and thermal-hydraulics coupling analysis. The DDC-3D code system couples the open-source codes DRAGON & DONJON based on two-step method and subchannel code COBRA-EN. The Picard iteration method is applied to ensure the stability of numerical calculation. Then the BEAVRS benchmark is used to validate the computational capabilities of DDC-3D code system. The critical boron concentrations, control rod worths and fission rate distributions are calculated in HZP condition. The results show a good agreement with measured data. The results demonstrate that the two-step method is applicable and valid for multiphysics simulations. For the result of HFP condition for cycle 1, the results also agree well with measured data, including the trend of the critical boron concentrations and power distributions throughout the cycle 1. Although the detailed thermal–hydraulic experimental values are not available, the thermal-hydraulics analysis of the hot fuel assemblies indicates that the calculation results are reasonable. In general, the results demonstrate the feasibility and accuracy of DDC-3D code system for neutronics and thermal-hydraulics coupling calculations and life cycle simulation of PWRs.

Development of CFD porous model parameters of Korean PWR spent nuclear fuel assemblies for thermal analysis

This study focused on developing and validating porous model parameters to accurately estimate the thermal behavior of spent nuclear fuel assemblies for computational fluid dynamics simulations. These assemblies are stored in Korean pressurized water reactors (PWRs) under dry storage conditions. Korean PWR fuel assemblies were categorized into four distinct types based on their configuration, facilitating the calculation of effective thermal conductivities and flow resistance coefficients. These parameters are crucial for simulating heat transfer and determining the peak cladding temperature (PCT), ensuring the thermal safety of stored nuclear fuel. The validation of the proposed porous model against explicit two-dimensional and three-dimensional simulations demonstrated its efficacy and reliability in enhancing the thermal analysis of spent fuel assemblies, highlighting the significance of accurately calculating the porous model parameters in the safety assessment of nuclear fuel storage. Furthermore, the study assessed the sensitivity of variables influencing temperature during spent nuclear fuel storage, focusing on flow resistance coefficients, control rod effect, and basket size. The findings revealed that variations in flow resistance and the presence of control rods minimally impacted PCT, whereas basket size significantly influenced temperature, underscoring its importance in the thermal analysis of nuclear fuel storage systems.

Operation Scheme analysis of a multipurpose small modular reactor under cogeneration condition based on a once-through steam generator dynamic model

Highly flexible load requirements and cogeneration economy need to challenge the operation of a multipurpose small modular reactor cogeneration plant with once-through steam generators. Therefore, the present study develops a once-through steam generator dynamic model to analyze the multipurpose small modular reactor operation scheme under cogeneration conditions at different power levels. The once-through steam generator dynamic model is derived based on conversation equations using the moving boundary method. It is verified with RELAP5 results and open literature and the maximum relative error is 3.21 %. Three operation schemes are proposed for multipurpose small modular reactor cogeneration operation: Scheme 1 with constant steam pressure and average coolant temperature, Scheme 2 with constant steam temperature and pressure, and Scheme 3 with constant steam temperature and pure sliding steam pressure. Steady-state and dynamic characteristics with three operation schemes are simulated and investigated at three power levels: 100 %, 70 % and 30 %. The steady-state results show that Scheme 1 is more favorable for the primary loop, while Scheme 2 is beneficial to the secondary loop, and Scheme 3 can improve the thermal efficiency at low power level. The transient findings indicate that disturbances from the reactor side have a significant impact on the once-through steam generator and the minimum settling time is 3.3 s. Consequently, steam temperature control of once-through steam generator is achieved by regulating the control rods for Schemes 2 and 3, while steam pressure is suggested to be controlled by the feedwater valve for Scheme 1. For cogeneration conditions at 70 % power level, Scheme 2 can achieve the highest steam flow of turbine, the highest steam flow of steam extraction and the smallest steam specific volume, which are 87.88 kg·s−1, 28.96 kg·s−1, and 0.0503 m3·kg−1, respectively. Scheme 2 is recommended for high power levels under cogeneration operation. In contrast, Scheme 1 is more suitable for the condensing unit operation for low power levels.

Reactor core design with enriched gadolinia burnable absorbers for soluble Boron-Free operation in the innovative SMR

Effective strategies for maintaining reactor reactivity at a consistent level are indispensable to enable soluble boron-free (SBF) operation in innovative small modular reactor (i-SMR) operation. This paper presents a reactor core design tailored for SBF operation in i-SMR, utilizing concentrated gadolinia burnable absorber rods. This investigation evaluates the concentration range of 155,157Gd and its reactivity characteristics, which are crucial for sustaining excess reactivity. The viability of employing enriched 155,157Gd for the i-SMR SBF operation is established through comprehensive analysis and simulations. Moreover, integrating 155,157Gd with integral gadolinia burnable absorber rods enhances the reactivity control and operational adaptability. Leveraging the spatial self-shielding effect of gadolinia extends the reactivity management capabilities and ensures the operational performance of the i-SMR as optimizing the cycle length. Core design considerations encompassing FA configurations and control rod patterns are investigated to ensure safety and performance across the initial and reload cycles. The feasibility of the SBF operation is verified by assessing compliance with the top-tier requirements of the i-SMR, thus underlining its potential for practical implementation.

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

The Russian fuel assembly TVS-2M is a newly proposed fuel configuration that contains gadolinium-oxide at different concentrations mixed with enriched UO2 at different U-235 enrichment values. The purpose of this study is to investigate the neutronic behavior of the Bushehr VVER-1000 reactor core by replacing the current standard Fuel Assemblies (FAs) called TVS with the new generation of TVS-2M FAs. In such manner, the reactor core was simulated under both modes of TVS-2M and TVS FAs arrangement using the WIMS, CITATION, MCNPX, and SuperMC codes in equal terms. Neutronic parameters of the reactor core such as effective multiplication factor, excess reactivity, neutron flux distribution, power peaking factor, and control rod worth are calculated and discussed within a comparative framework of both configurations of FAs; TVS and TVS-2M.

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.

COOLING OF CONTROL RODS AND SPECIFICATION OF CRITICAL TEMPERATURES TO JUSTIFY SAFE OPERATION

The paper analyzes the cooling conditions of the control rods in the mixed core of the WWER-1000 reactor containing fuel assemblies of different vendors (TVSA, WFA/RWFA). The coolant heat-up in the guide thimbles due to own energy release in control rod materials and control rod temperatures required to justify safe operation were calculated for the most conservative case, when the core consists only of TVSAs with a standardized guide thimble inlet plug of NCCP. The energy release was obtained using the calculation method for the boron carbide absorber (B4C), dysprosium titanate (Dy2O3·TiO2), control rod cladding (42CrNiMo), water gap between the cladding and the guide thimble and in the guide thimble (zirconium alloy), fully immersed and 70% immersed RCCA 10th control group. The highest calculated energy release will be observed in fully immersed control rods with boron carbide material at the height position corresponding to the zone of maximum neutron fluence and will be 43.8 W/cm3. The maximum energy release of the 70% immersed (from the bottom of the core) 10th control group is 29.4 W/cm3 for fresh B4C. For the condition of the reactor facility at which the temperatures in FA guide thimbles are maximum, for a "fresh" control rod characterized by maximum energy release in the neutron-absorbing material, the maximum coolant heat-up in the guide thimbles does not exceed 28.9 °C, with the maximum coolant heat-up in FAs being 44 °C, and the maximum control rod cladding temperature not exceeding 335.6 °C at a coolant saturation temperature of 345.8 °C, which indicates that there will be no coolant boiling in the guide thimbles under operation. The maximum temperature typical for the central part of the control rod absorber and for the most conservative calculation for a "fresh" absorber is 529.6 °C, which is significantly lower the melting point of boron carbide, which is ~2350 °C. The maximum calculated temperature in the contact zone of the B4C absorber with control rod cladding is 348 °C. The performed calculations and further analysis of the results confirm that during operation of the RCCA control rods with both “fresh” and “burned-up” B4C, there will be no boiling of the coolant and melting of the absorber in the central part of the control rod. This means that the VVER-1000 core with TVSAs and WFAs/RWFAs and the RCCA rods are reliably cooled.

Experimental validation of the 3-D neutron kinetics algorithm of RAPID using the JSI TRIGA Mark-II reactor

The high-fidelity simulation of reactor kinetics, with the ability of calculating detailed 3-D time-dependent distributions of the nuclear quantities of interest, is an important challenge in the field of nuclear engineering. As the industry moves towards innovative reactor designs and the financing and licensing of prototypes becomes increasingly complex, there is an increasing need of best-estimate and high-fidelity codes that are capable of replacing the computational tools that have been fine-tuned for LWRs using years of operation data and validation against benchmark facilities. Traditional neutron transport software either has to rely on significant modeling approximations or requires an impractical amount of computer time to calculate accurate and detailed solutions to the transport problems, failing to meet the need of the industry to simulate numerous scenarios over a relatively small time. RAPID is a software that relies on the Multi-stage Response-function Transport (MRT) methodology to achieve high-fidelity accurate solutions to the neutron transport problem by pre-calculating a database of response functions or coefficients that are then combined and interpolated to calculate the nuclear quantities of interest in seconds to minutes on a single computer core. In this article, the recently developed algorithms for 3-D time-dependent simulation of reactor kinetics and for control rods movement in the reactor core are validated using experiments performed at the Jožef Stefan Institute (JSI) TRIGA Mark-II reactor. The experiments were designed and performed with the goal of RAPID’s validation. In particular, time-dependent responses from four fission chambers located in the TRIGA core were collected and compared to RAPID-calculated values. The comparison of RAPID’s predictions to measured data shows very good agreement, demonstrating the ability of RAPID of calculating detailed 3-D time-dependent fission source distributions for a reactor kinetic problem that would be impractical to model with traditional neutron transport codes. In addition, it is demonstrated how RAPID is capable of obtaining a high level of accuracy within seconds to minutes on a single computer core, with a significant speedup with respect to state-of-the-art neutron transport methods.

Measurement and calculation of neutron flux deformation during rod drop transient at the VR-1 reactor

The rod drop experiment was performed at the VR-1 reactor to test the detection system and collect data that describe the development of neutron flux during rod drop. This experiment was calculated in Serpent code by its dynamic part. During this experiment, an additional control rod for drop was placed in the reactor with a channel directly attached to the rod guide tube containing the detection system. The detection system developed for this experiment is based on diamond detectors and allows measurement of neutron flux development during the rod drop in almost point vise location. The results of both experiment and calculation showed a delay in the count rate drop influenced by the position of the rod during the drop. The obtained experience is also used in the design process of a new experimental device dedicated to rod drop measurements that will allow the direct rod position measurement during the drop.

Detection of Axial Locations of Control Rods Using Movable In-core Detectors

Background Reactivity control in Pressurized Water Reactors (PWRs) relies on Rod Cluster Control Assemblies (RCCAs), or control rods (CRs), and boric acid diluted in primary circuit water. The control rods’ axial location inside the reactor core is typically provided by means of two redundant systems: the Rod Positioning Indicator (RPI) and the Rod Control System (RCS). Methods The current work focuses on the development of an additional method for the detection of control rod axial location inside the reactor core of a Pressurized Water Reactor (PWR) using the Advanced Flux Mapping System (AFMS) and its Movable In-core Detectors (MIDS). The paper presents the methodology employed and the results of the first validation campaigns. Results Two validation campaigns were carried out, targeting respectively equilibrium and non-equilibrium xenon axial distribution. The routine proved successful in identifying control rods’ axial insertion in different operating conditions including Flux Maps (FMs) acquired at 2%, 50% and 75% power. Conclusions The routine developed at Beznau Nuclear Power Plant will serve as an additional core monitoring tool to be employed in case of suspicious or faulty indication coming from the main control rod location methods. The developed method proved able to locate control rods’ axial insertion with few steps error, though there is room for potential improvements with regards to the uncertainty model currently implemented.