Rod Cluster Control Research Articles

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.

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.

Design and Development a Software for Wear Measurement of Control Rod Guide Card Based on Machine Vision

The control rod guide tube (CRGT) is used to protect the rod cluster control assembly (RCCA) from bending while providing guidance for the up and down movement of the RCCA. This ensures that the dropping time of the RCCA meets nuclear safety requirements and stops nuclear reactions within the specified time frame. During the up and down movement of the RCCA, the control rod guide cards (CRGC) may experience axial and vibration wear. If the wear is severe, the CRGC may lose its guiding function, resulting in serious problems such as delayed dropping time. In this paper, a machine vision method is proposed to measure the dimensions of the CRGC. Edge extraction, image rectification, and template matching are used to realize autonomous measurement of the CRGC based on its features. The developed measurement software has been used 5 times in domestic nuclear power plant overhauls and has detected 15 CRGTs that were close to the wear limit. This provided data support for the replacement of the CRGTs and effectively improved the quality of nuclear power operation.

Effects research of multiple-cycle nuclide-number-density distributions on dynamic rod worth measurement

The dynamic rod worth measurement (DRWM) has been widely applied in PWRs to measure the worth of RCCA (Rod Cluster Control Assembly) using the electricity signal of ex-core detector. In DRWM, the response function of ex-core detector is prerequisite for determining the spatial-correction factors of DRWM, which are vital parameters to correct the spatial effects due to the quick movement of RCCA during the measurement process. Theoretically, the response function of ex-core detector characterizes the contributions of the in-core neutron-flux distribution on the response of ex-core detector, which is dependent on the in-core nuclide-number-density distribution. Therefore, the researches have been implemented to quantify the effects of the multiple-cycle nuclide-number-density distribution on the response function of ex-core detector and further on the DRWM results. The previous microscopic-depletion model in our home-developed Bamboo-C code could only consider 29 main heavy-metal nuclides and 4 main fission-poison nuclides. In this research, the microscopic-depletion capability with full nuclides (233 nuclides) has been implemented, with which the detailed multiple-cycle nuclide-number-density distribution could be provided. For the method verification, two continuous refueling cycles of CNP1000 operated in China have been modeled, by which the effects of the multiple-cycle nuclide-number-density distribution on the response function of ex-core detector, then on the spatial-correction factors of DRWM and finally on the worth of RCCA have been quantified. Through the numerical results, it could be observed that the largest deviation of response function occurs at the fuel assembly near the ex-core detector, which is less than 3.0e-6 in absolute value. The relative deviations of spatial-correction factors are less than 1.5%, and the relative deviations of RCCA’s worth are less than 0.2%. Therefore, the 29 heavy-nuclide model is accurate sufficiently for the application of DRWM technique.

Numerical simulation of pressure force for control rod guide tube in PWR

In nuclear power plants (PWR), the rod cluster control assemblies (RCCAs) can be damaged by impacted-sliding wear due to flow induced vibrations, which leads to specific wear of the contact surface. A detailed numerical simulation of the flow field and pressure force in the control rod (CR) guide tube was carried out based on the CFD method in this paper. Then a new method was proposed to transfer the surface pressure filed to the pressure force on the control rods. The results show that the pressure force on the inner rods that located at the height of second guide card are higher than the others. At different positions of the guide cards, the pressure force of internal control rod CR1 is higher than CR2. The pressure force of the external control rod CR5 is lower than CR3 and CR4. For the same control rods, the pressure force at the height of guide card 2 is higher than guide card 1 and guide card 3. At the same time, the pressure force of control rod 1 is the highest than other rods that means the control rod 1 may result in larger potential displacement, on the contrary, the control rod 5 is likely to have a smaller potential displacement.

Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C

Based on the conventional “two-step” scheme for the PWR fuel management, an advanced PWR fuel management software Bamboo-C has been developed by the most advanced methodologies in the reactor-physics field. Bamboo-C consists of three main functional codes: LOCUST code for the heterogeneous modeling and simulation and homogenization calculation of 2D assemblies; SPARK code for 3D core steady-state and transient analysis; and LtoS code for assembly homogenization parameter function, which links LOCUST and SPARK. Bamboo-C has all the necessary analysis functions for the fuel management and nuclear design of PWRs, mainly including the start-up physics tests, calculations of the neutron-kinetics parameters, differential and integral worth of rod cluster control assemblies (RCCAs), and power-operation following simulation. Finally, the engineering validations of Bamboo-C have been completed according to the operation data from the reactors CNP300, CNP650 and CNP1000 designed by China independently. The validation results show that the errors between the values of such key parameters of cores as critical boron concentration, temperature coefficient, RCCA worth, and power distributions, calculated by Bamboo-C, and their measured values satisfy the corresponding engineering criterion limits.

Nuclear-data uncertainty analysis for the start-up physics test of CPR1000 reactor

In this paper, the uncertainty propagation has been performed from the nuclear data to the key parameters of start-up physics test for the first cycle of CPR1000 reactor. The start-up physics test is very important for the safety of PWRs, during which the key parameters including the critical boron concentration, temperature coefficient, integral worth of RCCA (Rod Cluster Control Assembly) and power distributions should be measured and used to validate the nuclear-designed ones. In order to guarantee the reactor safety, the errors between the simulation values and measurement values are required to satisfy corresponding engineering limits. However, for the conventional validation method, the uncertainties of the simulation values were ignored. Therefore, the uncertainty analysis has been performed to propagate the nuclear-data uncertainties to these key parameters of the start-up physics test for the CPR1000 reactor. Through the analysis, the absolute and relative uncertainties of these key parameters have been quantified, with the nuclear-data covariance data from ENDF/B-VII.1, ENDF/B-VIII.0 and TENDL-2017.

Electromagnetic Testing of Rod Cluster Control Assemblies in Pressured-Water Reactor Power Plants

This paper presents an electromagnetic testing system for rod cluster control assemblies used in pressurized-water reactors. The system uses several encircling-type magnetic cameras equivalent to a number of the control rods; each sensor probe composes of an encircling Hall sensor array (EHaS) and a bobbin coil. The EHaS has 16 Hall sensor elements that measure the electromagnetic field distribution in the radial direction of the control rod induced by the bobbin coil for defects. Experiments are performed on artificial defects on the cladding tube of real control rods to simulate short-circumferential grooves (SCGs), sliding wears (SWs), and circumferential cracks (CCs). The system can inspect the artificial SCGs, SWs, and CCs with depths up to 20%, 30%, and 40% of the cladding tube thickness (0.47 mm), respectively. Furthermore, the shape and depth of the defects could be estimated. The standard deviations of depth estimation are 18%, 5.8%, and 6.0% for CCs, SCGs, and SWs. The SCGs and SWs have a small and similar estimation error, but the CCs have the highest error of estimation, and have a small width of 0.2 mm.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Fretting wear of a nitrided 316L/304L contact subject to in-phase normal force fluctuation in dry and lithium-boron solution: An RP-friction energy wear approach

In nuclear power plants, tubes of the rod cluster control assemblies undergo impacts at low contact pressures against guides, causing specific wear on the contact surfaces. The present study investigates fretting wear behavior under variable loading, using an original double-actuator fretting wear system which allows independent control of tangential sliding and normal force fluctuation. Using this test system the fretting wear response of a nitrided 316L SS cylinder fretted against a 304L SS plate in air and in a lithium-boron solution is investigated.The effect of normal force fluctuations quantified using the RP=Pmin/Pmax ratio was investigated keeping constant the tangential sliding amplitude. Surface damage evolution was followed by 3D profilometry and several analyses were conducted on specimen surface and cross-sections (SEM, EDX). To quantify the wear volume extension, an RP-weight friction-energy-wear parameter (i.e.,RPη×ΣEd) is introduced. This new loading parameter allows us to take into account both friction work and normal force fluctuation.For a dry interface the η exponent was found to be very small (0.15) suggesting that the wear volume is directly related to the friction energy dissipated in the interface. In contrast, in lithium-boron solution the best fitting was achieved with ηw=−0.9 which implies an energy wear rate nearly inversely proportional to RP. The larger the normal force fluctuation (i.e., the smaller RP) the larger the contribution of tribocorrosion and the faster the surface wear. Besides, unexpected erosion-corrosion wear phenomena were identified at the inlets of the contact. These different wear mechanisms are discussed and quantified.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17×17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core retains design margin with respect to the MDNBR safety limit for both of the MSLB accident scenarios. The scenario with available offsite power was more restrictive in terms of MDNBR than the scenario without offsite power.

Fretting wear response of a nitrided 316L SS/304L SS interface: Effect of lithium/bore liquid environment

In nuclear power plants (NPP), rod cluster control assembly (RCCA) tubes undergo impact under low-pressure contact against guides, leading to specific wear on the contact surfaces. To focus on this industrial issue, an experimental layout was designed to perform reciprocating fretting in different environments.A representative tube interface (nitrided AISI 316L)/plane (AISI 304L) was investigated in air and in a solution composed of 1000ppm of bore and 130ppm of lithium to be closed to a NPP primary water chemistry at atmospheric pressure and room temperature. Wear regimes were identified in terms of variation in parameters such as sliding amplitude (from ±40μm to ±160μm), normal loads (from 2N/mm to 5N/mm) and test duration (from 100,000 to 1.5millioncycles).Surface damage evolution was followed by 3D profilometry and several analyses (SEM, EDX, optical observation) were conducted on surfaces and cross-sections. Based on these results, fretting-wear mechanisms in the 304L SS plate and nitrided 316L SS tube were investigated. The first results in dry environment showed that, for δg⁎ <±120μm, plane profiles had a “W-shape” whereas for δg⁎ >±120μm there was a U-shape. Thus, a W-shape morphology parameter Mw was defined, taking into account the third-body wear volume. Moreover, the worn surfaces of the plane were smoother in solution, and wear was no longer detected on the tube specimens. A modified energy wear approach was then implemented to quantify reciprocating fretting wear rate according to ambient conditions. Finally, two wear mechanisms are proposed.

Neutronics and thermal hydraulics analysis of a small modular reactor

The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO2) fuel assemblies; 5 assemblies at 4.4at% 235U enrichment and 64 assemblies at 4.95at% 235U enrichment. The active fuel length is 200cm and the core diameter is 194.55cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240cm resulting in an increased capacity factor of 95% at 530MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.

PWR control rods position monitoring

This study investigates the best radial positions for the monitoring detectors inside the reactor core, which provides close monitoring to enhance reactor safety. These positions depend on flux distortion due to reactivity change induced by accidental drop of a control rod into the core, as one of the causes of the reactivity change.Thermal neutron flux shape after each control rod group and rod cluster insertion is estimated to show the maximum distortion in neutron flux shape.The study aims to define the maximum distortion case by means of correlation coefficients corresponding to the data that were calculated using the three dimensional diffusion depletion code TRITON. Correlation coefficient and reactivity were determined for each rod cluster control (RCC) assembly. For each case, the change in shape of the thermal flux was determined in case of full insertion or accidental drop of one of the rod cluster control (RCC) assemblies in the reactor core at full power.

Improvement of core control strategy for CPR1000: Load follow without boron adjustment

This paper presents the study of load follow operations without boron adjustment for CPR1000. To enable the CPR1000 to perform load follow maneuvers without changing soluble boron concentration, the worth of Rod Cluster Control Assemblies (RCCAs) are reconfigured with their amount and location unchanged according to the reactivity variations during the load follow transient. To ensure the real-time ability of the reactor core, the target axial offset (AO) during load follow operations is set to the same value with that in based load, and the Delta-I is maintained within the special trapezoidal shaped target band around its target value.For the simulation of the reactor core, the time-dependent one-dimensional two group diffusion equations with the reactivity feedback of moderator temperature, Doppler and xenon–iodine are used. The transverse buckling is adjusted at each time interval so that the one-dimensional model can match the average characteristics of the three-dimensional reactor core accurately. To show the superiority of the improved core control strategy for CPR1000 reactor, the load follow results employing the purposed boron-adjustment free control strategy are compared to those obtained with the typical MODE-G control strategy. It has been demonstrated by the simulation results that the load follow capability of CPR1000 reactor is greatly improved due to the elimination of boron concentration adjustments during load change transients. Full load follow capability of the reactor has been extended from the initial 80% of cycle life to more than 90% of cycle life. Thus, the boron adjustment free improvement on the MODE-G core control strategy is feasible for CPR1000, which can improve the economical performance of the plant and simplify the operational process during power maneuvers.

Development of CFD methodology for investigating thermal-hydraulic characteristics in a PWR dome

This study aims to develop a detailed computational fluid dynamics (CFD) model to investigate the flow and heat transfer characteristics in the dome of a pressurized water reactor (PWR). The upper plenum is also considered in order to simulate the possible coolant leak to the dome via the gaps of upper support plate. The essential solid components within the solution domain, including the upper core plate, the guide tube assemblies, the support columns, and the rod cluster control, are realistically modeled, instead of the porous-medium approximation. Through the detailed-geometry CFD simulation, the thermal-hydraulic features in the upper plenum, individual guide tube assembly, and the dome can be obtained. And, the temperature distribution on the reactor pressure vessel (RPV) head can be used to estimate the values of total effective degradation years (EDY) and reinspection years (RIY) for monitoring the crack initiation and growth on the head. Present calculated results also reveal that the original values of EDY and RIY using the cold-leg temperature as the head temperature by the Maanshan staff is conservative.

Shutdown margin for high conversion BWRs operating in Th-233U fuel cycle

Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-233U fuel cycle (Th-RBWR). The studied core has an axially heterogeneous fuel assembly structure with a single fissile zone “sandwiched” between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Implementation of alternative reactivity control materials, reducing axial leakage through non-uniform enrichment distribution, use of burnable poisons, reducing number of pins as well as increasing pin diameter are also shown to be incapable of meeting the SDM requirements. Instead, an alternative assembly design, based on Rod Cluster Control Assembly with absorber rods was investigated. This design matches the reference ABWR core power and has adequate shutdown margin.The new concept was modeled as a single three-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules.

On the Use of Reduced-Moderation LWRs for Transuranic Isotope Burning in Thorium Fuel—I: Assembly Analysis

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which are thermal absorbers. These phenomena lead to increasing amounts of Pu feed being required to sustain criticality and accordingly larger TRU content in the multirecycled fuel inventory, ultimately resulting in a positive moderator temperature coefficient (MTC) and void reactivity coefficient. Because of the favorable impact fostered by use of thorium (Th) on these coefficients, the feasibility of Th-TRU multiple recycle in reduced-moderation pressurized water reactors (PWRs) and boiling water reactors (BWRs) has been investigated. In this paper, Part I of two companion papers, the analysis is limited to a single assembly, with full-core models presented in Part II. Spatial separation of TRU from bred uranium is found to greatly improve neutronic performance. A large reduction in moderation is necessary to allow full actinide recycle. This will pose thermal-hydraulic challenges, which are discussed in Part II. In addition, the harder neutron spectrum resulting from the reduced moderation also reduces the control rod worth, while there is a neutronic incentive to use increased mechanical shim to maintain a negative MTC. It may therefore be desirable to increase the number of rod cluster control assemblies. Superior burnup is achievable in a reduced-moderation BWR as a larger reduction in moderation is feasible, although the incineration rate is reduced relative to a PWR due to a higher conversion ratio.

DNBR margin analysis for uncontrolled withdrawal of rod cluster control assembly bank from subcritical or low power startup condition

DNBR margin analysis for uncontrolled withdrawal of rod cluster control assembly bank from subcritical or low power startup condition