Control Rod Guide Tubes Research Articles

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.

THEFIS Test Simulation to Validate a Freezing Model of ASTERIA-SFR Core Disruptive Accident Analysis Code

The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs.

A numerical investigation into Metallic-Melt continuous drainage in the core support plate region of a BWR for the initial phase of core melt progression

Under a typical uncovered core scenario in a Boiling Water Reactor (BWR), it is not clear that a large core molten pool will result as occurred in the Three Mile Island, Unit 2 (TMI-2) accident. Sandia National Laboratories (SNL) has propounded the alternative “Continuous Drainage” melt progression for the scenario, where molten material drains from the core region without the formation of any stable crusts or blockages. Hence the present simulation study was conducted to investigate the behavior of the metallic-melt continuous drainage and the drainage pathways during the initial phase of core melt progression in a BWR under an uncovered core condition like that encountered in the Fukushima Daiichi Nuclear Power Plants (NPPs) accident experienced by Tokyo Electric Power Company Holdings (TEPCO). Researchers around the world have conducted numerous sophisticated and comprehensive studies using simulation codes and experiments to investigate the Fukushima accident. Melt drainage with deforming components is subject to many uncertainties in a core support plate region possessing rather complicated geometries. Additionally, there is very limited experimental and numerical simulation knowledge on melt drainage in such a region. Thus, in recent years, Japan Atomic Energy Agency (JAEA) has experimented on melt drainage using a simplified apparatus to reproduce the behavior. The present paper provides simulation results on melt drainage in the core support region using SAGITTARIUS, an in-house three-dimensional simulation code capable of consistently simulating melt drainage with crust formation, local blockage, and structural damage. The in-vessel simulation target with a prototypic BWR-composition and -geometry is composed of four fuel assemblies (4 × 4 fuel rods per assembly), an absorber blade, a support piece, a core support plate, an Instrument Guide Tube (IGT), and a Control Rod Guide Tube (CRGT) with a Control Rod Drive Housing (CRDH). We estimated the drainage speed for the JAEA experiment to validate the SAGITTARIUS function. The drainage speed was evaluated to be 3.1 m/s, a speed corresponding to that observed in the core support plate region. We also used SAGITTARIUS to simulate the behavior of the metallic-melt drainage and temperature rise of the structure. The results from SAGITTARIUS on the trend and values of the temperature rise agreed with the measured values. On the basis of this finding, we conducted numerical investigations on the metallic-melt drainage of B4C-Stainless Steel (SS) eutectic from the absorber blade and Zry-melt from the fuel rods. The simulation demonstrated the capacity to describe the following features: (i) visual demonstration of metallic-melt drainage from the absorber blade and fuel rods, (ii) aggressive Zry-melt attack to the inlet orifice, (iii) crust formation with local blockage in the narrow drainage paths, and (iv) sedimentation of solidified melt on the bottom faces. These findings were consistent with the concept of the “Continuous Drainage” proposed at SNL and we supported the SNL scenario by the numerical investigation. These investigations probably provide useful knowledge during the discharge of damaged fuels or crust in the Fukushima-Reactor Pressure Vessels (RPVs).

CFD Simulation of thermal hydraulic characteristics in a typical upper plenum of RPV

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

Large eddy simulation of thermal mixing with conjugate heat transfer at BWR operating conditions

Thermal fatigue occurs in most metals under cyclic heat loads and can threaten the structural integrity of metal parts. Detailed knowledge of these loads is of utter importance to prevent such issues. In this study, a large eddy simulation (LES) with wall-adapting local eddy viscosity (WALE) subgrid model is performed to better understand turbulent thermal mixing in an annulus with a pair of opposing cold inlets at a low axial level (z=0.15m) and with a pair of opposing hot inlets at a higher axial level (z=0.80m). Each inlet pair is 90° from each other in the azimuthal direction. Conjugate heat transfer between fluid and structure is accounted for. The geometry simplifies a control-rod guide tube (CRGT) in a boiling water reactor (BWR). LES results are compared with measurement data. This is one of the first times BWR conditions are met in both experiments and LES: pressure equals 7.2 MPa, while the temperature difference between hot and cold inlets reaches 216 K. LES temperatures at the fluid-structure interface are fairly correlated with their experimental equivalents, with regard to mean values, local variances, and dangerous oscillation modes in fatigue-prone areas (z=0.65-0.67m). An elastic analysis of the structure is performed to evaluate stress intensities there. From them, cumulative fatigue usage factors (CUFs) are estimated and used as screening criteria in the subsequent frequency analysis of temperature time series at the fluid-structure interface. The likelihood of initiating a fatigue crack is linked to the maximum CUF, which is 3.2 × 10−5 for a simulation time of ~10s.

Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code

Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes

Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called “CRGT cooling”). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

Experimental investigation of the effect of control rod guide tubes on the breakup of a molten metal jet in the lower plenum of a boiling water reactor under isothermal conditions

It is important to clarify the molten material jet breakup process to estimate corium behavior in the lower plenum of a boiling water reactor (BWR). To identify the effect of control rod guide tubes (CRGTs) on the jet breakup behavior, a molten material (U-alloy) breakup experiment considering CRGTs in a BWR lower plenum was conducted under isothermal conditions. The experiment results show that jet breakup fraction for the case with CRGTs (pitch/diameter ratio of 1.37) was only approximately 20 % of that for the case without CRGTs. A coarser pitch/diameter ratio of 2.47 was also tested, but this configuration only slightly reduced the amount of jet breakup. The experiments also indicate that the CRGTs had no significant effect on the fragmentation of droplet diameter. Furthermore, the velocity distribution was measured around the jet with particle image velocimetry. The velocities of the water surrounding the jet for the cases with CRGTs were relatively larger than those in the case without CRGTs.

Severe accident progression in the BWR lower plenum and the modes of vessel failure

Most of our knowledge base on the severe accident progression in the lower plenum of LWRs is based on the data obtained from the TMI-2 accident. It should be recognized that the lower plenum of a BWR is very different from that of a PWR. Unlike the PWR, the BWR plenum is full of control rod guide tubes (CRGTs) with their axial structural variations. These CRGTs are arranged in a cellular fashion with each CRGT supporting 4 rod bundles. There are also a large number of instrument guide tubes (IGTs), each generally placed in the middle of 4CRGTs. Both the CRGTs and IGTs traverse the thick vessel bottom wall and are welded to their extensions which come to bottom of the core. The core-melt progression in the lower plenum is controlled by the structures present and they, in turn, influence the timings and the modes of vessel failure for a BWR.The uranium oxide–zirconium oxide core melt formed in the 4 fuel bundles is directed by the structure below toward the water regions in-between the 4 CRGTs. The FCI will take place in those water regions and some particulate debris will be created, although there is insufficient water for quenching the melt. A FCI may occur inside a CRGT if and when the melt enters the CRGT at its top opening or the melt in the water region between the four CRGTs breaches the wall of the CRGT.The important issue is whether the welding holding the IGT inside the vessel will fail and the bottom part of the IGT falls out creating a hole in the vessel with release of water and melt/particulate debris from the vessel to the dry well of the BWR containment. Similarly, the failure of CRGT could have water and melt/particulate debris coming out of the vessel. These modes of vessel failure appear to be credible and they could occur before any large-scale melting and melt pool convection takes place. These modes of vessel failure and the melt release to the containment will have very different consequences than those generated by the other modes of vessel failure.Such BWR plenum melt progression scenarios have been considered in this paper. Some results of analyses performed at KTH have been described. We believe that the issues raised are important enough to consider a set of experiments for verification and validation of the melt progression in a BWR plenum. Such experiments are proposed.

Heat transfer effect of an extended surface in downward-facing subcooled flow boiling

New BWR containment designs are considering cavity flooding as an accident management strategy. Unlike the PWR, the BWR has many Control Rod Guide Tube (CRGT) penetrations in the lower head. During a severe accident scenario with core melt in the lower plenum along with cavity flooding, the penetrations may affect the heat transfer on the ex-vessel surface and disrupt fluid flow during the boiling process. A small-scale experiment was performed to investigate the issues existing in downward-facing boiling phenomenon with an extended surface. The results were compared with a bare (flat) surface. The mass flux of 244kg/m2s, 215kg/m2s, and 177kg/m2s were applied in this study. CHF conditions were observed only for the 177kg/m2s case. The boiling curves for both types of surfaces and all flow rates were obtained. The boiling curves for the highest flow rate showed lower surface temperatures for the extended surface experiments when compared to the bare surface. The downstream location on the extended surface yielded the highest surface temperatures as the flow rate was reduced. The bubble accumulation and low velocity in the wake produced by flow around the extended surface was believed to have caused the elevated temperatures in the downstream location. Although an extended surface may enhance the overall heat transfer, a reduction in the local heat transfer was observed in the current experiments

The Effective Convectivity Model for Simulation of Molten Metal Layer Heat Transfer in a Boiling Water Reactor Lower Head

This paper is concerned with the development of approaches for assessment of core debris heat transfer and Control Rod Guide Tube (CRGT) cooling effectiveness in case of a Boiling Water Reactor (BWR) severe accident. We consider a hypothetical scenario with stratified (metal layer atop) melt pool in the lower plenum. Effective Convectivity Model (ECM) and Phase-Change ECM (PECM) are developed for the modeling of molten metal layer heat transfer. The PECM model takes into account reduced convection heat transfer in mushy zone and compositional convection that enables simulations of noneutectic binary mixture solidification and melting. The ECM and PECM are (i) validated against relevant experiments for both eutectic and noneutectic mixtures and (ii) benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is then applied to the analysis of heat transfer in a stratified heterogeneous debris pool taking into account CRGT cooling. The PECM simulation results show apparent efficacy of the CRGT cooling which can be utilized as Severe Accident Management (SAM) measure to protect the vessel wall from focusing effect caused by metallic layer.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper ( Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

Analytical Evaluation of Flow and Flow Induced Vibration Characteristics in the ABWR Lower Plenum

Reactor internal pumps (RIPs) are used in the advanced boiling water reactor (ABWR) nuclear power plants to circulate the reactor coolant in the reactor pressure vessel (RPV). Control rod drive housings (CRDHs) and control rod guide tubes (CRGTs) in the lower plenum were subjected to RIP discharge flow. Fluid force acting on those structures in the lower plenum must be evaluated in the next ABWR in which lower plenum structures are improved from the current design. Computational fluid dynamics (CFD) analysis is expected to evaluate the fluctuating fluid force acting on those structures. The 5 RIPs sector models of the reactor lower plenum were simulated to investigate the fluid force in the present study. Fluctuating fluid forces were calculated based on the velocity distribution obtained from the CFD analysis. Fluctuating fluid forces were applied to the structural models and those FIV stresses were calculated by the vibration analysis. As a result, flow in the lower plenum went along the RPV bottom and these flow characteristics were desirable for structures.

改良型沸騰水型軽水炉下部プレナム内の流動評価技術(NP2 軽水炉の設計・建設技術)

Reactor internal pumps (RIPs) are used in the advanced boiling water reactor (ABWR) nuclear power plants to circulate the reactor coolant in the reactor pressure vessel. Control rod drive housings (CRDHs) and control rod guide tubes (CRGTs) in the lower plenum were subjected to RIP discharge flow. Fluid force acting on those structures in the lower plenum must be evaluated in the next ABWR in which lower plenum structures have been improved from the current design. Computational fluid dynamics (CFD) analysis is expected to evaluate the fluid force acting on those structures. The 5 RIPs sector models of the reactor lower plenum were simulated to investigate the fluid force in the present study. Fluctuating fluid forces were calculated based on the velocity distribution obtained from the CFD analysis. Fluctuating fluid forces were applied to the structural models and its FIV stress was calculated by the vibration analysis. As a result, flow in the lower plenum went along the RPV bottom and this flow was desirable for structures. Remarkable fluid force acting on structures was not observed in the ABWR lower plenum.

Numerical simulation of the inlet plenum flow field for a LWR reactor pressure vessel

FLOW3D has been employed to simulate the turbulent flow occurring in the reactor pressure vessel (RPV) inlet plenum of the simplified boiling water reactor (SBWR) nuclear plant. Calculations have been carried out for a 90° sector of the RPV lower plenum, from the bottom to the core plate level, including downcomer and control rod guide tubes (CRGTs) simulation. A structured body-fitted grid is used to model geometrical details inside the plenum (CRGTs bundle and downcomer). Grid sizes of 162,000 nodes with the k-ε turbulence model have been used. In the present analysis, the usual porosity approach has not been used, while cross flow conditions through CRGTs bundle are taken into account by direct simulation of the tubes gap. Predictions of the steady-state pressure and velocity field have been made. The effects of the number of grid points and boundary conditions on the problem solution have been investigated. Results show that a CFD analysis is able to add new relevant information on the flow pattern inside the plenum, which is usually not taken into account by traditional engineering practice.