Control Rod Insertion Research Articles

Simulation of boiling water reactor one-pump trip transient by SIMULATE-3K

To validate models relevant to the transitional in-core 3-D power behavior, the 3-D core kinetics code SIMULATE-3K was applied to an analysis of one-pump trip test performed in a BWR-5 plant which includes large distortion of the power distribution due to the local insertion of control rods. The core boundary conditions required by SIMULATE-3K were calculated by the plant system analysis code RETRAN-3D. It was found that the transitional APRM signal and LPRM signals calculated by SIMULATE-3K agreed well with the measured data. The results showed that the transitional in-core 3-D power behavior can be appropriately predicted by SIMULATE-3K.

Development of a control rod depletion methodology for the Westinghouse NEXUS system

The NEXUS project is an effort to merge and modernize the methods employed in Westinghouse PWR and BWR steady-state reactor physics codes. The NEXUS system relies on a once-through nodal cross-section generation methodology with an innovative and efficient technique for pin power recovery. The pin power methodology overcomes a well-known limitation of existing methodologies, namely the incapacity to properly account for heterogeneity changes due to the depletion environment. The so-called control rod history problem where control rods are repeatedly inserted and withdrawn during core depletion is a good example of such a case. In addition to the control rod history impact on pin power distributions, the insertion of control rods during extended periods leads to significant control rod depletion that affects the reactivity worth of the control rods which in turn can have a significant impact on pin powers. The importance of accurately predicting pin powers, combined with the need to adequately estimate the reactivity worth and nuclear end of life of control rods in BWRs and in generation III+ PWRs, has motivated the development of a novel control rod depletion model. This methodology and its numerical qualification, initially for PWR application only, is the topic of this paper. The focus is on describing the salient features of the model and on illustrating its performance by means of numerical experiments. It is shown that together with the NEXUS pin power recovery model, the control rod depletion methodology accurately predicts the reactivity feedback from repeated control rod insertions in a PWR core.

Design simplification of a small nuclear reactor for large-diameter neutron transmutation doping silicon using control rods

A design concept for a small nuclear reactor for neutron transmutation doping silicon (NTD-Si) using a Pressurized Water Reactor (PWR) full-length fuel assembly was proposed in our previous work. The excess reactivity was suppressed by a combination of Gd2O3 and soluble boron, which results in a flatter flux profile over the core than with control rod insertion; however, the soluble boron system for reactivity control is quite complex and expensive. The removal of this system would make the design much simpler. In the current work, the removal of soluble boron is considered. Criticality, neutron transportation and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the insertion of control rods in five of the nine assemblies is enough to suppress reactivity. The thermal hydraulic analysis showed that heat removal from the core was possible under 1 atm operating pressure. Silicon ingots up to 30 cm in diameter could be irradiated with sufficient uniformity in the irradiation channels.

Safety implications of reactivity variations in fast thorium ADSRs

Nuclear power generation offers a reliable, low-impact and large-scale alternative to fossil fuels. However, concerns exist over the safety and sustainability of this method of power production, and it remains unpopular with some governments and pressure groups throughout the world. Fast thorium fuelled accelerator-driven sub-critical reactors (ADSRs) offer a possible route to providing further re-assurance regarding these concerns on account of their properties of enhanced safety through sub-critical operation combined with reduced actinide waste production from the thorium fuel source. The appropriate sub-critical margin at which these reactors should operate is the subject of continued debate. Commercial interests favour a small sub-critical margin in order to minimise the size of the accelerator needed for a given power output, whilst enhanced safety would be better satisfied through larger sub-critical margins to further minimise the possibility of a criticality excursion.Against this background, this paper examines some of the issues affecting reactor safety inherent within thorium fuel sources resulting from the essential Th90232→Th90233→Pa91233→U92233 breeding chain. Differences in the decay half-lives and fission and capture cross-sections of 233Pa and 233U can result in significant changes in the reactivity of the fuel following changes in the reactor power. Reactor operation is represented using a homogeneous lumped fast reactor model that can simulate the evolution of actinides and reactivity variations to first-order accuracy. The reactivity of the fuel is shown to increase significantly following a loss of power to the accelerator. Where the sub-critical operating margins are small this can result in a criticality excursion unless some form of additional intervention is made, for example through the insertion of control rods.

An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

Abstract Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Quantitative study of the possible operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven unique case studies are proposed based on a thorough review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. The qualitative parameters of each case study are outlined as well as the basis in literature. A previously published modeling scheme is described and adapted for application as a simulation platform for these transient events. The results of the quantitative case studies are described within part II of this paper.

An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

Abstract Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400MW design.

Effect of Polynomial Expansion Order of Intranode Flux Treatment in Nodal SN Transport Calculation Code NSHEX for Large-Size Fast Power Reactor Core Analysis

The nodal discrete ordinates (SN) transport calculation code for three-dimensional hexagonal geometry NSHEX treats intranode flux distribution using a polynomial series and considers the angular dependence of flux by the SN method. For the improvement of calculation accuracy of NSHEX for practical use to large-size fast reactor plants, the maximum order of the polynomial series is extended from two to six. In order to check the effect of the polynomial expansion order, NSHEX is applied to the intermediate-size fast power reactor core “Monju” and the large-size one “Super Phenix,” including various control rod insertion conditions. From the application, it is found that extension of the polynomial expansion order is effective especially for the large-size core “Super Phenix” under the control-rod-inserted condition.

Development of Dynamic Insertion Model of Control Rods of a Boiling Water Reactor under Seismic Events

In the seismic design of the boiling water reactor, we need to estimate the dynamic insertion behavior of control rods into the core region to secure the safety of the reactor under seismic events. In particular, estimation of the insertion time is one of the most important design tasks affecting the scrammability of the reactor. We developed a numerical analysis model to predict the control rod insertion time under seismic events using multibody dynamics. The effect of the interaction force between the control rod and the fuel assemblies is considered in three-dimensional contact analysis. This interaction force causes resistance force acting on the control rod under insertion. The hydraulic control unit and the control rod drive, which provide the control rod with driving force, were modeled in the concentrated parameter system considering dynamic characteristics, such as the inertance of the working fluid in the scram piping and the capacitance of the working gas in the accumulator. The numerical analysis can simulate the realistic insertion behavior of the control rod by coupling these models together and using an interactive process to calculate how they interact. The validity of the numerical analysis model was confirmed by comparing the analytical results with the experimental ones. The results of our analysis showed that the numerical analysis model provides good agreement with the insertion time of the control rod.

Effect of Polynomial Expansion Order of Intranode Flux Treatment in Nodal SN Transport Calculation Code NSHEX for Large-Size Fast Power Reactor Core Analysis

The nodal discrete ordinates (SN) transport calculation code for three-dimensional hexagonal geometry NSHEX treats intranode flux distribution using a polynomial series and considers the angular dependence of flux by the SN method. For the improvement of calculation accuracy of NSHEX for practical use to large-size fast reactor plants, the maximum order of the polynomial series is extended from two to six. In order to check the effect of the polynomial expansion order, NSHEX is applied to the intermediate-size fast power reactor core “Monju” and the large-size one “Super Phenix,” including various control rod insertion conditions. From the application, it is found that extension of the polynomial expansion order is effective especially for the large-size core “Super Phenix” under the control-rod-inserted condition.

Vessel coolant mixing effects on a PWR Main Steam Line Break transient

A Main Steam Line Break transient analysis for a Swiss operated Pressurized Water Reactor has been performed using the code RETRAN-3D, which included a 3-D core model applying a one-to-one correspondence between the thermal–hydraulic and the neutron kinetics nodalizations (nodal/channel) and dynamically coupled to a thermal–hydraulic plant system model. The analysis was postulated at Hot Zero Power conditions and was based on a real UO2-MOX core to whom conservative modifications were made to bring about a power excursion despite the control rods insertion. The objective of this work was to assess the impact on the core power evolution, at both the global and local scales, of the level of detail adopted for the description of the thermal–hydraulics upstream and downstream of the core. To that aim, a parameterized coolant mixing model in the Reactor Pressure Vessel was developed to control the level and the radial details of the coolant mixing at the inlet and outlet of the core. For the core considered in this work and using different credible coolant mixing patterns, it was observed that the coolant mixing in the Reactor Pressure Vessel could have different, and sometimes even opposite, effects depending on the scale that was considered and on the details of the assumed coolant mixing pattern.

Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR

In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 °C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs). Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated.

Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. Perturbation method is used to obtain the equation for the relative efficiency of control rod insertion. A series of coefficients, describing the axial absorption profile are used to correct the equation for a composite rod, having a complicated burn-up irradiation history. These coefficients have to be determined – by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross-sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn-up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct MCNPX evaluations of control rod worths is also presented.

Anomalous coolant temperature feedback effect in LWRs

In neutronically large cores of power reactors anomalous reactivity effects can occur due to space-dependent coolant temperature feedback. These effects are contrary to reactor physical intuition, e.g. power increase after control rod insertion. A quantitative analysis and physical interpretation of these anomalies is presented, based on two relatively simple reactor models and application of adjoint function theory.

Xenon-induced axial power oscillations in the 400 MW PBMR

The redistribution of the spatial xenon concentration in the 400 MW Pebble Bed Modular Reactor (PBMR) core has a non-linear, time-dependent feedback effect on the spatial power density during several types of operational transient events. Due to the inherent weak coupling that exists between the iodine and xenon formation and destruction rates, as well as the complicating effect of spatial variance in the thermal flux field, reactor cores have been analyzed for a number of decades for the occurrence and severity of xenon-induced axial power oscillations. Of specific importance is the degree of oscillation damping exhibited by the core during transients, which involves axial variations in the local power density. In this paper the TINTE reactor dynamics code is used to assess the stability of the current 400 MW PBMR core design with regard to axial xenon oscillations. The focus is mainly on the determination of the inherent xenon and power oscillation damping properties by utilizing a set of hypothetical control rod insertion transients at various power levels. The oscillation damping properties of two 100%–50%–100% load-follow transients, one of which includes the de-stabilizing axial effects of moving control rods, are also discussed in some detail. The study shows that, although first axial mode oscillations do occur in the 400 MW PBMR core, the inherent damping of these oscillations is high, and that none of the investigated load-follow transients resulted in diverging oscillations. It is also shown that the PBMR core exhibits no radial oscillation components for these xenon-induced axial power oscillations.

Analysis of Dynamic Insertion of Control Rod of BWR under Seismic Excitation

The dynamic characteristics of control rod for boiling water reactor being inserted under seismic excitation were investigated using non-linear analytical models. The capability of managing the insertion of control rod is one of the most important factors affecting the safety of nuclear power plant undergoing seismic events. Predicting the behavior of control rod being inserted during earthquakes is important when designing how rod should be controlled during seismic events. We developed analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in non-linear analysis. This interaction force causes resistance force to be applied to the control rod when they are being inserted. The validity of the analytical models was confirmed by comparing the analytical results with the experimental ones. The effects of input seismic motion and structural parameters on the insertion time ware investigated using the analytical models. These analytical methods can be used to predict the time to insert the control rod into the core region of reactor, and are useful for designing control rod system that can survive seismic events.

In-phase and out-of-phase modes stability analysis with LAPUR5 code for Kuosheng

A study of the in-phase (global, core-wide) and out-of-phase (regional) modes of stability with LAPUR5 code for Kuosheng NPP is presented. Comparisons made with vendor’s STAIF results indicated close agreements, within the acceptable 20% deviation in decay ratios, for Kuosheng NPP Unit 1 Cycle 18 reloads design. Parametric studies for stability have shown that the inlet loss coefficient, exposure, control rod position, and axial power shape are the most important parameters that might affect the accuracy of decay ratios for in-phase and out-of-phase modes of stability analysis. A decrease in the inlet loss coefficient or increase in the control rods insertion causes an increase in the global and regional decay ratios. An increase in the exposure causes an increase in the global decay ratio, the maximum value of the regional decay ratio occurs in MOC condition, and the regional decay ratio is higher than global decay ratio in the BOC condition.

Development of core fuel management code system for WWER-type reactors

In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

Feasibility Study for Evaluation of Control Rod Worth in Pressurized Water Reactors using Neutron Count Rate during a Control Rod Drop Testing

Applicability of the modified Neutron Source Multiplication (NSM) method with extraction of the fundamental mode to subcriticality measurement has been proposed. Following the feasibility verification in the previous study based on numerical analyses, its applicability has been proven in a more realistic situation; in a withdrawal sequence of control rod banks during the PWR startup. Subcriticalities with various control rod insertion configurations were estimated based on the modified NSM method. The subcriticality could be evaluated with a good accuracy even with the mockup experiment where any special treatments for accurate measurement were not taken into account and furthermore the insensitivity of measured signals by reactivity changes and their large fluctuations were seen. Based on this fact, we further investigated a feasibility to use neutron count rate data obtained during the control rod drop testing, which is carried out before the reactor physics tests at hot zero power condition. When it is proven that these data could be used for the estimation of each control rod worth, the following reactor physics tests could be performed with the advanced knowledge of each control rod worth and procedures for detailed control rod worth measurement could be simplified or eliminated from the reactor physics tests.

06/01613 Analysis of NEA C5MOX benchmark with computer code ‘COHINT’ based on interface current approach: Mohanakrishnan, P. Progress in Nuclear Energy, 2004, 45, (2–4), 125–131

The 3-D extension C5G7 MOX benchmark, which is a follow-up benchmark of the C5G7 MOX on deterministic 2-D/3-D MOX fuel assembly transport calculations without spatial homogenization, aims at providing a more challenging test of the abilities of currently available 3-D methods to handle the spatial heterogeneities of reactor core. The geometry configuration was modified with some significant changes including control rod insertion in order to check real 3-D effects with the previous benchmark. In this work, THREEDANT code within DANTSYS code package, which solves the multi-group discrete ordinates form of the Boltzmann transport equation in x–y–z and r–z–theta geometries, was employed to perform the benchmark calculations with the spatial approximation to a square shape of the circular fuel pin and the angular approximation to S8. The calculation results from THREEDANT indicate that the power decrements due to control rod insertion are larger than those in the reference solution and are also more sensitive. In addition, a sensitivity study on mesh refinement was performed to produce a higher quality solution. In the results, it is found that the 4 × 4 grid per pin cell in the THREEDANT calculation is sufficiently refined so that very little benefit is obtained by increasing the mesh size.