Reactivity Insertion Transients Research Articles

Development of a 3D APOLLO3® Neutron Deterministic Calculation Scheme for the CABRI Experimental Reactor

CABRI is an experimental reactor at the French Alternative Energies and Atomic Energy Commission (CEA) used to study fuel behavior during reactivity insertion transients. As these transients have a high level of multiphysics, it is important to develop suitable modeling and simulation tools to simulate them in order to be able to optimize testing and control of experimental conditions. This paper focuses on the development and validation of the neutron deterministic APOLLO3® calculation scheme that is included in the CABRI neutronic/thermal-hydraulic multiphysics coupled simulation tool; it represents the first stage of a stepwise validation process for the CABRI multiphysics simulation tool. The neutron calculation scheme is based on a classical two-step approach. The first step consists of a 281-energy-group flux calculation with the TDT-MOC (Method of Characteristics) solver for nuclear data space and energy collapsing for the different CABRI assembly clusters. The biases on a two-dimensional (2D) core neutron calculation due to self-shielding correction and collapsing on a restricted pattern are investigated by means of comparison with a direct full 2D calculation on a quarter core. The second step relies on a three-dimensional (3D) core calculation. Two approaches are presented. The first one consists of a best-effort approach corresponding to a 3D pin-scale description of the core, performing a transport calculation with the SN solver MINARET. And, the second one, a best-estimate approach, which will be implemented for kinetics calculations, relies on solving a simplified transport SPN equation in the solver MINOS with an exact 3D cell description of the core. The best-estimate calculation scheme is then used to analyze three experimental CABRI transients. A stepwise validation process is followed to quantify the calculation biases on physical parameters such as reactivity, reaction rates, and total core power at each step using static reference calculations with the stochastic code TRIPOLI4® or transient experimental data. The next development stage toward a multiphysics calculation scheme will be implementation and validation of coupling with a core thermal-hydraulic model.

MNSR transient analysis using the RELAP5/Mod3.2 code

To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 °C at 240 s, 124 °C at 160 s and 126 °C at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.

Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

Abstract This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK-1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

Impact of fuel thermal conductivity degradation on Doppler feedback during rod ejection accident

This paper discusses the importance of the fuel thermal conductivity degradation modeling for accurate predictions of the Doppler feedback during reactivity insertion transients. The impact of the fuel thermal conductivity degradation model, recently implemented in the coupled sub-channel thermal-hydraulic/time-dependent neutron transport code system CTF/TORT-TD, on Doppler feedback predictions during a control rod ejection accident was investigated. The rod ejection was simulated for a 4×4 pressurized water reactor pin array, extracted from the Purdue University MOX (mixed oxide) benchmark, starting at both hot zero power and hot full power conditions with the control rod being half-inserted before the ejection. The two scenarios were simulated with CTF/TORT-TD and the effect of the fuel thermal conductivity degradation on the Doppler feedback was analyzed. The results were compared with existing reference calculations performed with the coupled sub-channel thermal-hydraulic/time-dependent neutron transport/fuel performance code system CTF/TORT-TD/FRAPCON-FRAPTRAN.The power pulse, the time evolution of average fuel temperature, and the peak enthalpy rise during the transient were examined. It was confirmed that the impact of the fuel thermal conductivity degradation is more significant when the control rod is ejected at hot full power conditions. If the fuel conductivity degradation was not taken into account, less conservative CTF/TORT-TD predictions for the transient power response were obtained. For the selected 4×4 pin array, the coupled code calculated 13MW higher power pulse when modeling degradation effects on fuel conductivity. The difference in the power response is due to the less negative prompt fuel temperature (Doppler) coefficient at elevated temperatures. Lower thermal conductivity will lead to higher fuel pellet temperatures, and subsequently to a less negative Doppler coefficient, which will result in a stronger power pulse. The maximum fuel enthalpy rise during the hot full power rod ejection accident was found to be 60cal/g (251,208J/kg).

Superheated Water-Cooled Small Modular Underwater Reactor Concept

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at 500oC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

Abstract The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

Effect of high density dispersion fuels on transient behavior of MTR type research reactor under multiple reactivity transients

The effect of fuel on transient behavior under simultaneous multiple reactivity insertion in a typical MTR type research reactor have been studied. The PARET code was used for three different high density fueled cores. The fuels included oxide (U3O8-Al), silicide (U3Si2-Al) and aluminide (UAlx-Al) respectively. In first part single ramp reactivity insertions were done and results were found in good agreement with already published data. Then simultaneous multiple reactivity insertion transients were investigated with totally disabled control systems. Multiple reactivities were inserted in the IAEA benchmarked MTR type research reactor and effects on reactor power and temperatures (fuel, clad and coolant) were studied. The simultaneous accidents included ramp reactivity (large and small) plus beam tube flooding, a ramp reactivity plus a core movement and a ramp reactivity plus a sample ejection accident respectively. In all cases, the reactivity rises to peak values and then starts decreasing slowly; however when the ramp is large the peak values also become high and power also shows multiple peaks. The core having oxide fuel has slightly higher power peak values while core with silicide and aluminide fuels show similar behavior. The fuel temperatures in all cases were much higher for the core having oxide fuel as compared to other two cores.

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.

Simultaneous multiple reactivity insertions in a typical MTR-type research reactor having U3Si2–Al fuel

Safety assessment for simultaneous multiple reactivity insertions in a typical research reactor having low enriched, high density (U3Si2–Al) fuel has been done using modified PARET code. As a first step, the effect of single reactivity ramps on power and temperatures during transients showed good agreement with already published data. Then simulation of transients for multiple reactivity insertions were carried out and the effects on reactor power, net reactivity of the system, and temperatures (fuel, clad and coolant) were studied. The simultaneous reactivity insertions included ramp reactivity (large and small) and beam tube flooding, a ramp reactivity and a core movement and a ramp reactivity and a sample ejection accident respectively. When simultaneously two different reactivity insertion transients are coupled in the reactor under uncontrolled conditions, the self-control of the reactor decreases. The power and temperatures show several peaks and it oscillates along with the net reactivity of the system within the first second of the accident.

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.

A model for nuclear research reactor dynamics

In this paper, a useful thirty-fourth order model is presented to simulate the kinetics and dynamics of a research reactor core. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant and fuel temperatures (Doppler effects) with variable reactivity coefficients, xenon, samarium, boron concentration, fuel burn up and thermal hydraulics. WIMS and CITVAP codes are used to extract neutron cross sections and calculate the initial neuron flux respectively. The purpose is to present a model with results similar to reality as much as possible with reducing common simplifications in reactor modeling to be used in different analyses such as reactor control, functional reliability and safety. The model predictions are qualified by comparing with experimental data, detailed simulations of reactivity insertion transients, and steady state for Tehran research reactor reported in the literature and satisfactory results have been obtained.

A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition.

Thermal–hydraulic modeling of reactivity insertion in a research reactor

Transient reactivity insertion for a pool type research reactor with scram enabled is numerically investigated. The analyses were performed by developing a coupled kinetic–thermal–hydraulics code, with continuous feedback due to coolant and fuel temperature effects. A modified Runge–Kutta method was adopted for a better solution to a system of stiff point kinetic equations. Transient thermal–hydraulics during the process of slow and fast reactivity insertion were considered for a IAEA 10-MW research reactor. The code was used to carry out transient thermal–hydraulics analysis for both high-enriched uranium (HEU) and low enriched uranium (LEU) fuel. Reactor power, fuel, clad and coolant temperatures, MDNBR, reactor period, energy released for slow and fast reactivity insertion transients were predicted. The model prediction was verified through comparison with the PARET code and other institution results reported in the literature for detailed simulations of reactivity insertion. The comparison of simulation results and the PARET code showed good agreement for the main thermal–hydraulic parameters.

Determination of thermal hydraulic data of GHARR-1 under reactivity insertion transients using the PARET/ANL code

The PARET/ANL code has been adapted by the IAEA for testing transient behaviour in research reactors since it provides a coupled thermal hydrodynamic and point kinetics capability for estimating thermal hydraulic margins. A two-channel power peaking profile of the Ghana Research Reactor-1 (GHARR-1) has been developed for the PARET/ANL (Version 7.3; 2007) using the Monte Carlo N-Particle code (MCNP) to determine the thermal hydraulic data for reactivity insertion transients in the range of 2.0 × 10−3 Δk/k to 5.5 × 10−3 Δk/k. Peak clad and coolant temperatures ranged from 59.18 °C to 112.36 °C and 42.95 °C to 79.42 °C respectively. Calculated safety margins (DNBR) satisfied the MNSR thermal hydraulic design criteria for which no boiling occurs in the reactor core. The generated thermal hydraulic data demonstrated a high inherent safety feature of GHARR-1 for which the high negative reactivity feedback of the moderator limits power excursion and consequently the escalation of the clad temperature.

Measured and simulated reactivity insertion transients characteristics of NIRR-1

The power excursion characteristics of the Nigeria Research Reactor-1 (NIRR-1) under different reactivity insertion transients have been calculated using PARET/ANL 7.3 code. Experimental data of dynamic experiments performed in NIRR-1 facility during initial startup were used to benchmark the calculated data. For ramp insertion of total cold core excess reactivity of 3.77 mK, the reactor quickly reaches a maximum power of 82.3 kW in 390 s and decreases gradually due to the inherent safety features. The calculated maximum power for this insertion is in good agreement with a measured value of 79.4 kW recorded during the commissioning of NIRR-1. Similarly, moderator temperature for the hottest channel with respect to this transient was calculated to be 63.7 °C which is comparable to a measured value of 63.9 °C. Furthermore, step reactivity insertions of 2.23 mK and an approximately 0.32 mK, a regular criticality experimental check carried out at MNSR facilities by Safeguards Inspectors were accurately simulated.

Reactivity feedbacks of a high-density low enriched uranium fuelled material test research reactor during reactivity insertion transients

The reactivity feedback of a high-density low enriched uranium fuelled material test research reactor during reactivity insertion transients was studied. For this purpose, the low density LEU fuel of an MTR was replaced with U-Mo (9w/o) high-density LEU fuels currently being developed under the RERTR program. Simulations were carried out to determine the reactivity feedback during reactivity insertion transients. The nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that during the fast reactivity insertion transient, which was also a super prompt critical transient, the reactor power depended upon the Λ / β eff factor and despite the fact that the feedback reactivities increased with the increase in uranium density, the reactor power increased as the fuel uranium density increased. During the slow reactivity insertion transient, the effect of increased Doppler feedback coefficient was visible. Although the maximum reactor power achieved during the transient increased slightly as uranium density increased, the energy released decreased. The net total reactivity decreased as uranium density increased and the net reactivity of the highest density fuelled reactors remained below the 1$ mark.

Uncontrolled reactivity insertion transients of a material test research reactor fuelled with high density U 3Si 2 dispersion fuels

The effects of using high density low enriched U 3Si 2 uranium dispersion fuels on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U 3Si 2 LEU fuels currently being developed under the RERTR program. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained followed the relationship: P ( M W ) = 8.256 + 6.276 ρ U − 0.714 ρ U 2 + 26.667 × 10 − 3 ρ U 3 for 4.0 ≤ ρ U ( g / cm 3 ) ≤ 6.0 When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point. The time of ONB followed the relation: t ONB ( s ) = 359.4 × 10 − 3 + 4.4 × 10 − 3 ρ U + 687.5 × 10 − 3 ρ U 2 for 4.0 ≤ ρ U ( g / cm 3 ) ≤ 6.0 The magnitude of total feedback reactivity at ONB decreased as uranium density increased and followed the relation: ρ fb ( $ ) = 130.0 × 10 − 3 + 19.3 × 10 − 3 ρ U + 2.4197 ρ U 2 for 4.0 ≤ ρ U ( g / cm 3 ) ≤ 6.0

Prospects of using different clad materials in a material test research reactor – part 4 – the uncontrolled reactivity insertion transients

The effects of using different clad materials on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the aluminium clad of an MTR was replaced separately with stainless steel-316 and zircaloy–4. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, all the reactors stabilised at new power levels. The SS cladded fuel attained the minimum power level of 17.72 MW, followed by Zr-4 cladded with 23.62 MW while it was maximum at 24.31 MW for Al cladded fuel. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the Al cladded was the first to reach ONB while SS cladded was the last.

Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm 3, 7.74 gU/cm 3 and 8.57 gU/cm 3. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm 3 to 8.90 gU/cm 3. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

TINTE analysis of PBMR reactivity insertion transients

Several important PBMR reactivity insertion transients have been simulated by means of the dynamic reactor code TINTE (Time-dependent Neutronics and Temperatures). These transients include total control rod removal during normal operating conditions, cold zero power (CZP) conditions and hot zero power (HZP) conditions, as well as Reserve Shut-down System (RSS) removal during the cold zero power conditions. According to the TINTE results, the worst control rod removal scenario is the total control rod removal during hot zero power conditions in a postulated event with the power conversion unit (PCU) starting up and with continual operation without a trip. The amount of reactivity insertion by the RSS removal is much greater than the reactivity insertion by control rod removal. The event was analyzed but the RSS during the cold zero power conditions is a Beyond Design Basis Accident (BDBA). This paper presents the methodology followed to model these transient events with the TINTE code, as well as total power, maximum and average fuel temperatures and case comparison results. It is also shown that the maximum fuel temperatures for all of the Design Basis Accidents (DBA) are within acceptable safety limits.