Rod Ejection Accident Research Articles

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.

REA 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS v2.7 and SIMTAB cross-sections tables

The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA at Hot Zero Power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS v2.7 a REA in Almaraz NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions.

ICONE19-44013 ROD EJECTION ACCIDENT BY THE COUPLED SYSTEM CODE ATHLET-QUABOX/CUBBOX

The paper considers a Rod Ejection Accident (REA) which has been calculated by the coupled-code system ATHLET-QUABOX/CUBBOX. For the present study, a MOX/UOX mixed core loading was developed on the basis of a generic PWR. The results are particularly focused on the fuel enthalpy rise which is the main safety criterion for such transient. A parametric REA study has been performed, showing the influence of some important thermal-hydraulic and neutron-physical parameters. Simulations have been performed using realistic or artificially decreased delayed neutron fractions for two different core states (HZP and 30% of the nominal power). Effective fuel rod temperature influence (i.e. Doppler coefficient) has been studied by using different correlations (0.5/0.5 weighting factors or the typical T_<Doppler> = 0.7 T_<Surface> + 0.3 T_<Center>) or by changing the fuel gap conductance. It is shown that the maximum enthalpy (and enthalpy increase) does not always appear in the affected fuel assembly but can also appear in the neighboring ones. This result is a direct consequence of the burn up dependence of the enthalpy. The paper also considers the case of local delayed neutron parameters and briefly describes the future REA studies foreseen at GRS such as an investigation of quantitative uncertainty propagation from the nuclear data to the transient behavior.

An Analytical Criterion to Prevent PCMI Fuel Rod Cladding Failure during RIA Transients

This paper describes a new approach to define an analytical fuel rod cladding failure criterion during fast power transients such as reactivity initiated accidents (RIAs). Though cladding failure is not a safety issue in itself in such situations, a criterion aiming at precluding this risk is a useful decoupling limit allowing to meet the safety issues. Based on experimental results, a failure limit, expressed in critical strain energy density (CSED) as a function of the waterside corrosion level or rod average burnup, has been determined. The CSED failure limit is then transposed for PWR conditions in terms of critical energy deposition (DHc) in the fuel. This is done by the means of CYRANO3 (steadystate) and SCANAIR (transient) calculations using relevant input data (rod design, power history, RIA transient shape, etc.). The DHc limit is a function of both rod average burnup and initial linear power. It is applicable to UO<sub>2<sub/>/Zy-4 rods submitted to rod ejection accidents (REAs), during which the cladding is subject to PCMI mechanical loading. It also precludes the risk of failure by oxidation and embrittlement at low burnup. The robustness of the DHc failure criterion is ensured by the penalties and uncertainties taken into account at the different steps of its elaboration

Study of the PWR REA pulse width for realistic UO2 and MOX core designs using 3D kinetics methods

For the development of new acceptance criteria for the analysis of Rod-Ejection-Accidents (REA) in Pressurized Water Reactor UO2 cores, full-width-at-half-maximum power pulse widths in the range 25–40 ms were employed in the analytical transient fuel behaviour studies. Due to the operation of MOX cores in Switzerland, a study using the CORETRAN 3D kinetics code was carried out at PSI to investigate and compare on the basis of real operated cycles MOX and UO2 cores, the pulse width magnitude as function of rod reactivity for postulated super-prompt critical REAs. This study is presented in a first part of this paper. Thereafter, to quantify the variation in pulse width when applying a different 3D kinetic solver, a benchmarking with the SIMULATE-3K code is carried out. This code-to-code comparison aims also to serve as an assessment of SIMULATE-3K for REA analyses of the Swiss cores since only CORETRAN had so far been benchmarked at PSI for this type of applications. Finally, to provide confidence in the results, particularly for pulse widths below 25 ms, sensitivity studies are performed to assess the effects of specific modelling options and assumptions applied in the REA 3D kinetic analysis and to compare the sensitivities between UO2 and MOX cores. On the basis of all these studies, it is found that the MOX core pulse width is usually around 5 to 10 ms lower than for UO2 cores. For rod worths close to 2.0$, corresponding to around 500 pcm above prompt criticality and which was obtained for many of the investigated MOX cycles particularly at end-of-cycle conditions, a representative MOX pulse width magnitude is found to be around 15 ms. To shortly address the applicability of these results, i.e. obtained for very specific MOX cores, additional calculations were performed for hypothetical cases with increased MOX core fractions and with higher MOX assembly Pu enrichments. And it was found that even for these types of cores, the pulse width would, although being slightly reduced, remain in the range 10–15 ms for rod reactivities up to 500 pcm above prompt criticality.

Application of Monte Carlo Method to Solve the Neutron Kinetics Equation for a Subcritical Assembly

There is a need to understand the space-dependent kinetics of fast or thermal reactor physics and the Monte Carlo method should be implemented in kinetics codes as well. In a transient accident (for example, control rod ejection accident or loss of coolant accident), changes in the system are much slower than the prompt neutron lifetime. In the present paper, a method is proposed in which the Monte Carlo method is used to solve the neutron kinetics of reactors, such as the case of a subcritical static reactor with an external neutron source. In the proposed method the time derivative of the neutron equation is set to zero, though the time derivative of the delayed neutron precursor equation is treated without any approximation in the time differentiation of delayed neutron precursor equation. Most kinetics calculation methods originate from the point kinetics method, and they require calculation of excess reactivity at each time step, but the present method does not need calculation of excess reactivity. The neutron counts measured with small fission chambers in the subcritical kinetics experiment using the JAEA fast critical assembly are compared with counts calculated by the proposed method and compared with the measured values. The agreement is good, and the excellent performance of the method for space-dependent neutron kinetics is shown.

Development of System Response Analysis Method for Rod Ejection Accident of OPR1000 and APR1400 Using RETRAN Code

The Korea Electric Power Research Institute has developed the non-loss-of-coolant accident analysis methodology, KNAP, for the typical Optimized Power Reactor 1000 and Advanced Power Reactor 1400 in Korea. The RETRAN hot spot model (HSM) has been developed to replace the functions of STRIKIN-II code of ABB-CE. To estimate the feasibility of HSM, the typical cases of rod ejection accidents (REA) were analyzed and the results were compared with those mentioned in Final Safety Analysis Report of Ul-chin Units 3 & 4 or Standard Safety Analysis Report of APR1400. Through this study, it was concluded that the HSM of KNAP could be applicable to the object plants from the viewpoint of REA.

Development of System Response Analysis Method for Rod Ejection Accident of OPR1000 and APR1400 Using RETRAN Code

The Korea Electric Power Research Institute has developed the non-loss-of-coolant accident analysis methodology, KNAP, for the typical Optimized Power Reactor 1000 and Advanced Power Reactor 1400 in Korea. The RETRAN hot spot model (HSM) has been developed to replace the functions of STRIKIN-II code of ABB-CE. To estimate the feasibility of HSM, the typical cases of rod ejection accidents (REA) were analyzed and the results were compared with those mentioned in Final Safety Analysis Report of Ul-chin Units 3 & 4 or Standard Safety Analysis Report of APR1400. Through this study, it was concluded that the HSM of KNAP could be applicable to the object plants from the viewpoint of REA.

Safety Analysis of High-Power-Density Annular Fuel for PWRs

This paper assesses the performance of internally and externally cooled annular fuel in a four-loop pressurized water reactor during a variety of transients and accidents, namely, the loss of flow accident (LOFA), main steam line break (MSLB), large break loss of coolant accident (LBLOCA), and rod ejection accident (REA). The RELAP5 code was the primary vehicle for these analyses, although the VIPRE code was also used to calculate the minimum departure from nucleate boiling ratio (MDNBR) for LOFA and MSLB transients based on the RELAP5 results. It has been found that the MDNBR for the annular fuel at 150% power was higher than the MDNBR value for the reference solid fuel at 100% power for LOFA and MSLB. For LBLOCA analysis, the RELAP5-3D code was applied twice since the code has a constraint on the reflood model, which can be applied to only one cooling surface (either the inner channel or the outer channel). The analysis, with the reflood model applied to the outer channel, showed that using the standard size (100%) accumulator but with an increased (150%) safety injection flow rate, the peak cladding temperature (PCT) for the annular fuel at 150% power would be ~1200 K (927°C). This is ~150°C higher than the PCT for the solid fuel at 100% power but 277°C lower than the regulatory limit of 1204°C. When the reflood model is applied to the inner channel, the PCT would be limited to 1100 K (827°C), which is only 50°C higher than the PCT for the solid fuel at 100% power and 377°C lower than the regulatory limit of 1204°C. The calculated fuel temperatures and enthalpies during the REA have been found to be much smaller for the annular fuel, even at 150% power, compared to that for the solid fuel at 100% power. These analyses indicate that the new internally and externally cooled annular fuel can accommodate 50% power uprate in a PWR and still maintain adequate safety margins for a variety of transients and accidents including LOFA, MSLB, LBLOCA, and REA.

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis

The rod ejection accident (REA) and the main steam line break (MSLB) are two of the most important design basis accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal-hydraulics codes for reactivity insertion accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that to perform cross code comparative analysis for REA and MSLB scenarios. The coupled 3-D neutron kinetics/thermal-hydraulics benchmark presented in this paper is based on data from the Unit #6 of the Bulgarian Kozloduy Nuclear Power Plant (KNPP) and it is entitled the VVER-1000 coolant transient (V1000CT) benchmark. Two real plant transients are selected for simulation in the benchmark: main coolant pump start-up (Phase 1) and coolant mixing tests (Phase 2). In addition to these transients extreme scenarios were defined for better testing 3-D neutronics/thermal-hydraulics coupling: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP (Phase 1) and MSLB transient (Phase 2). The paper presents an overview of the Phase 1 (V1000CT-1) benchmark activities and describes the approach used for assessing the coupled neutron kinetics/thermal-hydraulics codes. Selected comparative analysis of currently submitted participants' results is presented with emphasis on the observed modeling issues and deviations from the measured data.

Computational Assessment of Burnup-Dependent Fuel Failure Thresholds for Reactivity Initiated Accidents

Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds associated with clad tube failure and fuel incipient melting are determined with respect to fuel burnup in the range from 30 to 70MWd(kgU) 1. The thresholds for clad tube failure are calculated by means of a strain-based clad failure criterion, which rests upon data from more than 200 out-of-pile mechanical property tests on recrystallized Zircaloy-2 and stress relieved annealed Zircaloy-4 cladding. The enthalpy thresholds for fuel incipient melting are calculated, considering the depression of fuel melting point with increasing burnup as well as burnup-related changes to the pellet radial temperature distribution. The calculated enthalpy thresholds for both clad tube failure and fuel incipient melting drop moderately with increasing burnup, and the thresholds are similar for the pressurised and boiling water reactor accidents under consideration. Extensive sensitivity studies are carried out, in order to quantify uncertainties in the calculated bestestimate thresholds.

Computational Assessment of Burnup-Dependent Fuel Failure Thresholds for Reactivity Initiated Accidents

Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds associated with clad tube failure and fuel incipient melting are determined with respect to fuel burnup in the range from 30 to 70MWd(kgU) 1. The thresholds for clad tube failure are calculated by means of a strain-based clad failure criterion, which rests upon data from more than 200 out-of-pile mechanical property tests on recrystallized Zircaloy-2 and stress relieved annealed Zircaloy-4 cladding. The enthalpy thresholds for fuel incipient melting are calculated, considering the depression of fuel melting point with increasing burnup as well as burnup-related changes to the pellet radial temperature distribution. The calculated enthalpy thresholds for both clad tube failure and fuel incipient melting drop moderately with increasing burnup, and the thresholds are similar for the pressurised and boiling water reactor accidents under consideration. Extensive sensitivity studies are carried out, in order to quantify uncertainties in the calculated bestestimate thresholds.

Development of Pressurized Water Reactor Integrated Safety Analysis Methodology Using Multilevel Coupling Algorithm

The subchannel code COBRA-TF has been introduced for an evaluation of thermal margins on the local pin-by-pin level in a pressurized water reactor. The coupling of COBRA-TF with TRAC-PF1/NEM is performed by providing from TRAC to COBRA-TF axial and radial thermal-hydraulic boundary conditions and relative pin-power profiles, obtained with the pin power reconstruction model of the nodal expansion method (NEM). An efficient algorithm for coupling of the subchannel code COBRA-TF with TRAC-PF1/NEM in the parallel virtual machine environment was developed addressing the issues of time synchronization, data exchange, spatial overlays, and coupled convergence. Local feedback modeling on the pin level was implemented into COBRA-TF, which enabled updating the local form functions and the recalculation of the pin powers in TRAC-PF1/NEM after obtaining the local feedback parameters. The coupled TRAC-PF1/NEM/COBRA-TF code system was tested on the rod ejection accident and main steam line break benchmark problems. In both problems, the local results are closer than before the introduced multilevel coupling to the corresponding critical limits. This fact indicates that the assembly average results tend to underestimate the accident consequences in terms of local safety margins. The capability of local safety evaluation, performed simultaneously (online) with coupled global three-dimensional neutron kinetics/thermal-hydraulic calculations, is introduced and tested. The obtained results demonstrate the importance of the current work.

Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis

A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near guide tubes, corners, and other sections where excessive power would create significant loss to thermal-hydraulic margins.This work examined a number of physics and core safety analysis parameters that impact the operation and safety of power reactors. Such parameters as moderator coefficients of reactivity, Doppler coefficients, soluble boron worth, control rod worth, prompt neutron lifetime, and delayed-neutron fractions were considered. These in turn were used to examine reactor behavior during a number of operational conditions, transients, and accidents. Such conditions as shutdown from power with one rod stuck out, steam-line break accident, feedwater line break, loss of coolant flow, locked rotor accidents, control rod ejection accidents, and anticipated transients without scram (ATWSs) were examined.The analysis of selected reactor transients demonstrated that it is feasible to license and safely operate a reactor fueled with plutonium-thorium blended fuel. In most cases analyzed, the thorium mixture had less-severe consequences than those for a core comprising low-enriched uranium fuel. In the analyzed cases where the consequences were more severe, they were still within acceptable limits. The ATWS accident condition requires more analysis.

The coupled kinetic and thermal-hydraulic three dimensional code system NLSANMT/COBRA-IV for PWR core transient analysis

A three dimensional neutron kinetic nodal code NLSANMT has been developed based on the semi-analytical nodal expansion method and nonlinear iteration strategy, and it has been coupled with the subchannel thermal hydraulic core analysis code COBRA-IV to form the integrated code system NLSANMT/COBRA-IV for PWR core transient analysis. This paper describes the features and status of the integrated NLSANMT/COBRA-IV system. A demonstration application of the NLSANMT/COBRA-IV code to rod ejection accident analysis is presented using the OECD NEACRP PWR hot zero power full core rod ejection benchmark problem. The NLSANMT/COBRA-IV solution agrees well with the reference solution, even when one node per assembly and one channel per assembly are used.

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor

A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching at room temperature. The test showed that the latching force increased with the current of the magnet coil, as did the de-latch time. The post-test analysis with a finite element analysis code revealed that the clearance between the two shafts greatly affected the latching force. With the same analysis method, the design analysis of the latch mechanism at a high temperature condition of 300°C was conducted, and it was confirmed that the latch mechanism contained enough latching force.

Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement.

Intercomparison of results for a PWR rod ejection accident

This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the United States, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general, the agreement between methods was very good, providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

Development of In-vessel Type Control Rod Drive Mechanism for Marine Reactor

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water—310°C and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch the driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature water for a long period. Under condition of high temperature water, the performance tests on latching, de-latching, holding and vertical movement, and durability test were conducted to verify the design conditions. This CRDM can be applied to not only the marine reactor, but also to light water reactors, PWR and BWR.

Development of In-vessel Type Control Rod Drive Mechanism for Marine Reactor.

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water—310°C and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch the driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been developed to be capable of working under the high temperature water for a long period. Under condition of high temperature water, the performance tests on latching, de-latching, holding and vertical movement, and durability test were conducted to verify the design conditions. This CRDM can be applied to not only the marine reactor, but also to light water reactors, PWR and BWR.