Rod Withdrawal Research Articles

The ATWS analysis of one control rod withdraw out of the HTR-10GT core in addition with bypass valve failure

The 10MW high temperature gas cooled test reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basic accidents of HTR10-GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism, in spite of the failure of bypass valve. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1283.3°C. It was a litter than 1230°C – the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation will be raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM for they have the same outlet temperature of the core.

STATUS OF THE ASTRID CORE AT THE END OF THE PRE-CONCEPTUAL DESIGN PHASE 1

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a ‘large pin/small spacing wire’ concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

Interpretation of the Control Rod Withdrawal Test in the Sodium-Cooled Fast Reactor Phénix

Before the definitive shutdown of the prototype Phénix, a final set of experiments was performed to gather important data about the operation and safety of sodium-cooled fast reactors (SFRs).Among the accident sequences that are to be taken into account, inadvertent withdrawal of a control rod is considered. During operation at nominal power, such a sequence induces a general power increase and local deformations of the power shape. Afterward, local fuel temperature increases can thereby lead to fuel melting and clad failure.The quasi-static control rod withdrawal test was specially designed to gather local power data on fissile assemblies and to complete validation databases of neutronic codes. The maximal deformation of the power shape reached ±12% and was obtained when two control rods were shifted in opposite directions.The test analysis was conducted with the neutronics code ERANOS-2.2. Comparisons between calculated and measured values were satisfying. Most of the discrepancies in power estimation can be explained by measurement problems (heat transfer, sodium mixing).The association of ERANOS-2.2 and the nuclear library JEFF-3.1, presently used for the predesign phase of the ASTRID reactor, constitutes an acceptable predictive tool for local and integral parameter estimations in SFRs, specifically in the evaluation of the control rod withdrawal incident.

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58gU/cm3) and enrichment (93%) of HEU fuel (U–Al), which may result in the changes of heat transfer response and neutronic characteristic in the core. So it is necessary to objectively re-assess the feasibility of LEU silicide fuel core in KUR by using various numerical simulation codes. This paper established a detailed simulation model for the LEU silicide core and provided the safety analyses for the reactivity insertion transients in the core by using EUREKA-2/RR code. Although the EUREKA-2/RR code is a proven and trusted code, its validity was further confirmed by the comparison with the predictions from another two thermal hydraulic codes, COOLOD-N2 and THYDE-W at steady state operation. The steady state simulation also verified the feasibility of KUR to be operated at rated thermal power of 5MW. In view of the core loading patterns, the operational conditions and characteristics of the reactor protection system in KUR, the accidental control rod withdrawal transients at natural circulation and forced circulation modes, the cold water injection induced reactivity insertion transient and the reactivity insertion transient due to removal of irradiation samples were conservatively analyzed and their transient characteristic parameters such as core power, fuel temperature, cladding temperature, primary coolant temperature and departure from nucleate boiling ratio (DNBR) due to the different ways and magnitudes of reactivity insertions were focused in this study. The analytical results indicate that the quick power excursions initiated by the reactivity insertion can be safely suppressed by the reactor protection system of KUR in various initial power levels and different operational modes (natural circulation and forced circulation modes). No boiling and no burnout on fuel cladding surface and no blister in the fuel meat happens and KUR is safe in all of these reactivity insertion transients if the reactor protection system of KUR works in its minimum degree.

International benchmark on the natural convection test in Phenix reactor

The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs.

Design of instrumentation for control and safety rod drive mechanisms of prototype fast breeder reactor

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, plutonium oxide-uranium oxide fuelled, pool type fast breeder nuclear reactor under construction in India. Control and Safety Rod Drive Mechanism (CSRDM) is part of one of the shutdown systems and is used to drive neutron absorber rods up and down. Instrumentation and control for CSRDM realises significant functions like power control, power set back and controlled shutdown. A study was carried out into incidents during the operation of 40 MWt experimental reactor called Fast Breeder Test Reactor and testing of a prototype CSRDM. Significant incidents in FBTR like uncontrolled withdrawal of absorber rod, movement in unintended direction, partial absorber rod drops are found to have occurred. Corrective measures are taken in the I&C of PFBR. PFBR I&C for CSRDM is presented. I&C designers for nuclear systems may gainfully make use of the experiences described in the paper.

Coupled code calculation of rod withdrawal at power accident

The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.

Thermal hydraulic simulation of reactor of HTR-PM based on thermal-fluid network and SIMPLE algorithm

This paper provides the model development and its verifications for the reactor thermal-hydraulic transient model for the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). A thermal-fluid network is constructed to simulate the complexity of the flow and heat transfer structure in the reactor of HTR-PM. SIMPLE algorithm is applied to solve the conservation equations of the thermal-fluid network to simulate the transient behavior of the high speed turbulent helium flow. A calculation process is proposed for coupling the high speed helium flow and the complex heat transfer structure. A FORTRAN code was developed based on the solution method and the thermal-fluid network. Several test cases including two steady states and a Control Rod Withdrawal accident are simulated by this code and the results are compared to those obtained by a safety analysis code, namely THERMIX. The good agreement between the two codes indicates that the proposed model and solution method based on SIMPLE algorithm is reasonable and applicable for simulating the thermal-hydraulic behavior in reactor of HTR-PM.

Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG

Abstract The multi-group version of the DYN3D reactor dynamics code was used for calculations for a new concept of a boiling water reactor with tight lattice of fuel rods and reduced neutron moderation. For that purpose, a 5-group cross section library was prepared and connected to the DYN3D code. Comparison calculations with the steady-state finite-difference code ACADEM showed a very good agreement. The capability of the DYN3D multi-group code in modeling transients in boiling water reactors with tight fuel element lattices was demonstrated by the analysis of two reactivity accidents initiated by the ejection of one control rod and unauthorized withdrawal of a control rod bank from the reactor core. The corresponding analyses were performed for begin of cycle conditions, when the considered control rods are at their maximum insertion depth.

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO2 fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.

Numerical analysis of startup PWR with fractional neutron point kinetic equation

In this paper we present the numerical analysis of the neutron density behavior when the nuclear reactor power is increased during startup of a PWR. The fractional neutron point kinetic (FNPK) equation with one-group delayed neutron precursor and external neutron source was used for this analysis. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and this effect is considered with the FNPK which have a physical interpretation of the fractional order is related with the sub-diffusive process, i.e., non-Fickian effects from the neutron diffusion equation point of view. In order to study the relaxation time effects during start-up of a PWR, a numerical analysis with FNPK is carried out, which it is assumed that during the ith step of control rod withdrawal the way of reactivity insertion is step to step, where the neutron source strength was defined as a constant in terms of a known initial stable sub-criticality and the neutron signal from a steady state condition. The results of the FNPK were compared with the classical neutron point kinetics (CNPK), for different values of the anomalous relaxation time.

A comparative study of FARCOB and ERANOS-2.1 neutronics codes in predicting the Phenix control rod withdrawal end-of-life experimental results

Abstract The post-analysis of control rod withdrawal end-of-life experiments carried out in the Phenix reactor during June 2009 is carried out at IGCAR by using the FARCOB and ERANOS 2.1 neutronics code systems. The measured results and the geometrical data required for this study is provided by the IAEA as a part of IAEA-CRP on “Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHENIX End-of-Life Experiments”. Four critical core states have been simulated with the average descriptions of the core. Parameters estimated are absorber rod worth, radial power distribution and the deviation in sodium heating change with respect to the reference state. All the predictions are based on 3-D diffusion theory calculations. This is the first experimental validation of radial power tilt due to control rod movement at full power with FARCOB code system. This study also enabled the benchmarking of calculation procedures and code systems being used at IGCAR for applications to oxide fueled sodium cooled fast power reactors. It is found that there is a good agreement in the predicted results from FARCOB and ERANOS 2.1 systems for this test.

Validation of 3D kinetics code 'TRIKIN' using OECD PWR core transient benchmark

A space-time kinetics code, ‘TRIKIN’, was developed and validated to analyse a vast array of transient problems involving reactivity and power distribution anomalies in VVER reactors. Neutronics model of TRIKIN is based on centre mesh finite difference approach for solving space- and time-dependent neutron flux in a 3D core whereas thermal hydraulics model is based on fuel pin simulation and employs a semi-implicit scheme for numerical solution. The code was validated against a series of rod ejection accidents with different levels of reactivity feedback. With likely induction of square lattice Pressurised Water Reactors (PWRs) in India, capabilities of TRIKIN code have been extended to analyse space-time kinetics related problems in this geometry. Newly added features of TRIKIN have been validated against an NEA rod withdrawal accident benchmark problem at Hot Zero Power (HZP) state. It has been demonstrated that TRIKIN results are in agreement with benchmark results.

APPLICATION OF BACKWARD DIFFERENTIATION FORMULA TO SPATIAL REACTOR KINETICS CALCULATION WITH ADAPTIVE TIME STEP CONTROL

The backward differentiation formula (BDF) method is applied to a three-dimensional reactor kinetics calculation for efficient yet accurate transient analysis with adaptive time step control. The coarse mesh finite difference (CMFD) formulation is used for an efficient implementation of the BDF method that does not require excessive memory to store old information from previous time steps. An iterative scheme to update the nodal coupling coefficients through higher order local nodal solutions is established in order to make it possible to store only node average fluxes of the previous five time points. An adaptive time step control method is derived using two order solutions, the fifth and the fourth order BDF solutions, which provide an estimate of the solution error at the current time point. The performance of the BDF- and CMFD-based spatial kinetics calculation and the adaptive time step control scheme is examined with the NEACRP control rod ejection and rod withdrawal benchmark problems. The accuracy is first assessed by comparing the BDF-based results with those of the Crank-Nicholson method with an exponential transform. The effectiveness of the adaptive time step control is then assessed in terms of the possible computing time reduction in producing sufficiently accurate solutions that meet the desired solution fidelity.

Effects of fuel type on the safety characteristics of a sodium cooled fast reactor. Part II: Simulation results

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In Part I of this paper, the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core were introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA ( RELAP5-3D, 2009) code are also presented and discussed. In this Part II of our paper, we present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios. Loss of Flow, Loss of Heat Sink, Loss of Power and inadvertent control rod withdrawal accidents were simulated for each core at beginning (BOC), middle (MOC) and end of a fuel cycle (EOC). The simulations were stopped at the initiation of melting of fuel or cladding. The results showed that in most of the transients the metal core came closer to its melting temperature while the strong reactivity feedbacks of the oxide and nitride cores limited their fuel temperature increases. Overall, the oxide and nitride cores had similar performance with respect to their inherent safety characteristics.

Investigation of axial power gradients near a control rod tip

Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15–3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Transient behaviour of 4.8 g/cc silicide fuel-(I) abnormal withdrawal of control rod during a start-up of JMTR

The Japan Materials Testing Reactor (JMTR) loaded the 4.8 g/cc silicide fuel for the core conversion from the high to the low enrichment (<20 wt%). In the experiment, the reactivity insertion of 0.35%Δk/k/s caused by the abnormal withdrawal of the control rod at the start-up of the JMTR was assumed. The value was about twice the EUREKA-2 computer code used for licensing. As a result, the test specimen (4.8 g/cc silicide fuel) had the peak cladding surface temperature of 137 ± 3°C. The test specimen did not fail by quench mechanism because the time to quench ( t q) was 0.11 s and the temperature drop Δ T was 28°C. Usually the quench failure occurs at the rapid t q (<0.13 s) and the high Δ T (>94°C). The safety margin of the JMTR against the accident induced by the abnormal withdrawal of control rod was assured.

Reactivity Accident in a High Temperature Gas-Cooled Reactor Due to Inadvertent Withdrawal of Control Rod

Abstract Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Development of TSAC1.0 and application to reactivity insertion accident of CARR

In this study, we have developed a thermo-hydraulic and safety analysis code named TSAC1.0 with Visual Fortran 6.5 to analyze the thermal-hydraulic characteristics of the China advanced research reactor (CARR) under reactivity insertion accident (RIA) which was induced by unexpected control rod withdrawal in full power condition. The neutron kinetic model depended on the point kinetics with six groups of delayed neutrons including reactivity feedback effects and it was adopted for the solution of reactor power. Furthermore, a new simple and convenient model was adopted for the solution of the transient behaviors of main pump instead of the complicated four-quadrant model. Visual input, real-time processing and dynamic visualization output were achieved using Microsoft Visual Studio.NET 2003 to make the application of TSAC1.0 much more convenient in the engineering. The simulated results of TSAC1.0 were found to be in reasonable agreement with those of RELAP5/MOD3 and showed that the parameters, including the peak coolant temperature, the peak heat structure temperature, and MDNBR, were in the acceptable range of design safety limit under RIA.

Validation of the DALTON-THERMIX Code System with Transient Analyses of the HTR-10 and Application to the PBMR

The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor–10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.