Safety Analysis Of Research Reactor Research Articles

Research Reactor Safety Analysis under Partial Loss of Flow without SCRAM

Research Reactor Safety Analysis under Partial Loss of Flow without SCRAM

Extension and validation of the SubChanFlow code for the thermo-hydraulic analysis of MTR cores with plate-type fuel assemblies

Many research reactors with plate-type fuel are operated worldwide and the thermal hydraulic core analysis is done with dedicated codes or with 1D system thermal hydraulic codes. At the Karlsruhe Institute of Technology (KIT), investigations are underway to develop advanced coupled neutronic/thermal hydraulic simulation tools for the Safety analysis of research reactors with plate-type fuel. For this purpose, the subchannel thermal hydraulic code SubChanFlow was extended and validated using relevant tests. The SubChanFlow extension consists of the addition of a heat conduction module for a thin plate and the implementation of heat transfer correlations for rectangular channels. For the validation of the modified SubChanFlow code, temperature data from RA-6 experiment obtained in a Reynolds range of 7×103<Re<1.4×104 were used.The RA-6 experiment was carried out at the Bariloche Atomic Center to support the development of MTR research reactors. Different heat transfer correlations e.g., the Colburn, Dittus-Boelter, Gnielinski, Sieder-Tate, and Y-Sudo were used to simulate the RA-6 tests. The comparison of the plate temperatures measured along the active height with the predicted values using the extended SubChanFlow code shows a deviation ranging from ± 4 °C to ± 18 °C for the Colburn and Y-Sudo correlations, respectively, which is promising compared to results of other codes.

Benchmark study of OPAL research reactor using MCNP codes

Abstract This work is a part of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) titled: \Benchmarks of Computational Tools against Experimental Data on Fuel Burnup and Material Activation for Utilization, Operation and Safety Analysis of Research Reactors". The OPAL research reactor has been modelled using MCNP codes as a benchmark study. Neutron flux profiles, critical configurations, control rod worths, fuel burnup and kinetic parameters have been calculated. Comparisons between the calculated results and the experimental data given in the OPAL benchmark have been performed.

Monte Carlo and nodal neutron physics calculations of the IAEA MTR benchmark using Serpent/DYN3D code system

As part of recent efforts to utilize NPPs computational methodologies to safety analysis of research reactors, the Serpent and DYN3D codes were extensively compared with a variety of static and burnup calculations as defined in the IAEA benchmark for 10 MW MTR pool-type reactor. These calculations include unit cell calculations and few group constants generation, unit cell and full core k-eigenvalue and burnup calculations, and full core 3D flux and power distributions. The Serpent code capabilities as a lattice code for MTR plate-type fuel assemblies were evaluated and compared with EPRI-CELL and WIMS-D4 results and reference solutions for full 3D core models were compared with MCNP5 and OpenMC results. The DYN3D nodal diffusion code capabilities in modeling full 3D MTR cores were also evaluated using few group cross sections and assembly discontinuity factors obtained by Serpent unit cell calculations. The DYN3D results were compared with Serpent, MCNP5 and OpenMC.

Thermal hydraulic analysis of ETRR-2 using RELAP5 code

Abstract The present work was developed within the frame of the IAEA Coordinated Research Project 1496 “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.

Application of best estimate plus uncertainty in review of research reactor safety analysis

To construct and operate a nuclear research reactor, the licensee is required to obtain the authorization from the regulatory body. One of the tasks of the regulatory authority is to verify that the safety analysis fulfils safety requirements. Historically, the compliance with safety requirements was assessed using a deterministic approach and conservative assumptions. This provides sufficient safety margins with respect to the licensing limits on boundary and operational conditions. Conservative assumptions were introduced into safety analysis to account for the uncertainty associated with lack of knowledge. With the introduction of best estimate computational tools, safety analyses are usually carried out using the best estimate approach. Results of such analyses can be accepted by the regulatory authority only if appropriate uncertainty evaluation is carried out. Best estimate computer codes are capable of providing more realistic information on the status of the plant, allowing the prediction of real safety margins. The best estimate plus uncertainty approach has proven to be reliable and viable of supplying realistic results if all conditions are carefully followed. This paper, therefore, presents this concept and its possible application to research reactor safety analysis. The aim of the paper is to investigate the unprotected loss-of-flow transients "core blockage" of a miniature neutron source research reactor by applying best estimate plus uncertainty methodology. The results of our calculations show that the temperatures in the core are within the safety limits and do not pose any significant threat to the reactor, as far as the melting of the cladding is concerned. The work also discusses the methodology of the best estimate plus uncertainty approach when applied to the safety analysis of research reactors for licensing purposes.

Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis.The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety analysis codes that comprise of CATHARE, RELAP5, MERSAT and PARET. The code RELAP5 was used independently by four of the participating teams and therefore the user effect and its impact on the code results can be characterized. The benchmark results demonstrate that most of the codes have the capability to correctly predict the SS case. However, for the LOFA case the simulation results show discrepancies to the measurement although the majority of the applied codes predict a qualitative correct time evolution of the corresponding transients for the coolant and clad temperatures. It is noted that the peak temperatures and the gradients around them are predicted conservatively. The quantitative assessments of benchmark results indicate different amounts of discrepancy between predictions and measurements ranging between 7% and 20% for peak clad temperatures during LOFA. The comparative prediction capability of the employed codes is addressed by additional code-to-code comparisons based on selected TH parameters that comprise flow rate, pressure drop and heat transfer coefficient during natural circulation phase.

Benchmarking RSG-GAS reactor thermal hydraulic data using RELAP5 code

The present work is developed within the frame of the IAEA Coordinated Research Program 1496, “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”.The objective of this work is to test the credibility of RELAP5 code in simulating the research reactor behavior during transients. Where, the code’s results were compared to the experimental measurements taken from Instrumented (thermocouple) Fuel Elements (IFE), installed at several positions in the core.The research reactor used in this study is the RSG-GAS reactor, located in the Serpong in Indonesia. Both the steady state and the transient simulation results performed using RELAP5 code for, are presented in this work.The calculated values for the coolant and clad surface temperature, at different locations in the core, showed an agreement with the experimental values with a difference less than 7% for the steady state, and with a difference less than 10% for transient conditions. However, RELAP5 was unable to predict the measured value of the coolant output temperature after the natural convection begins especially after stagnation occurred. In this case, the difference between the code results and experimental data was about 23%.

Investigation of the Capability of REALP5 to Solve Complex Fuel Geometry

This work is developed within IAEA Coordinated Research Program 1496, “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. The study investigates the capability of Code RELAP5/Mod3.4 to solve complex geometry complexity. Its results are compared to the results of PARET, a common code in thermal hydraulic analysis for research reactors, belonging to MTR-PC groups. The WWR-SM reactor at the Institute of Nuclear Physics (INP) in the Republic of Uzbekistan is simulated using both PARET and RELAP5 at steady state. Results from the two codes are compared. REALP5 code succeeded in solving the complex fuel geometry. The PARET code needed some calculations to obtain the final result. Although the final results from the PARET are more accurate, the small differences in both results makes using RELAP5 code recommended in case of complex fuel assemblies. Keywords—Complex fuel geometry, PARET, RELAP5, WWRSM reactor.

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed. In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage. For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.

NUR research reactor safety analysis study for long time natural convection (NC) operation mode

The current work represents a part of an overall effort being undertaken to improve NUR nuclear research reactor neutron utilization capabilities using a new core configuration. A RELAP5 model for the reactor NUR under natural convection (NC) operating mode has been developed. The model represents internal pool reactor components with the corresponding geometry, point neutron kinetics, and thermal hydraulics characteristics. An experimental device was designed and implemented in the reactor pool for monitoring the inlet and outlet core temperatures, and other pool temperature positions during NC-operating mode. In this paper, unprotected fast reactivity insertion and total flow blockage of the flapper valve transients have been investigated under NC operating conditions. The achieved steady-state results were found to be in good qualitative agreement with measurements. The results obtained from the transient (FRIA) study were compared to similar approach from recent literature. The second transient herein considered, is an attempt to predict the reactor core thermal-hydraulic behavior under a total flow blockage of the (NC) flapper valve. The latter could be considered as useful contribution for updating the safety analysis report.

Modification and validation of the natural heat convection and subcooled void formation models in the code PARET

Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions.

Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor

RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

Simulation of subcooled flow instability for high flux research reactors using the extended code ATHLET

Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the Gesellschaft for Anlagen- und Reaktorsicherheit (GRS-Society for Plant and Reactor Safety) (Lerchel, G., Austregesilo, H., 1998. ATHLET Mode 1.2 Cycle A, User's Manual, GRS-p-1/Vol. 1, Rev. 1, GRS). In order to extend the code's range of application to the safety analysis of research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime (Hainoun, A., 1994. Modellierung des unterkühlten Siedens in ATHLET und Anwendung in wassergekühlten Forschungsreaktoren, D 294 Diss. Univ. Bochum, Jül-2961). Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended program, the Thermal–Hydraulic Test Loop (THTL) of Oak Ridge National Laboratory (ORNL) was modeled with ATHLET and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET-postcalculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW m −2 and inlet flow velocity up to 20 m s −1. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12 and 7%, respectively.

Modelling of void formation in the subcooled boiling regime in the athlet code to simulate flow instability for research reactors

The athlet thermohydraulic code was developed at the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic non-equilibrium effects in the subcooled boiling regime. The aim of the extension is, on one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors owing to their high power densities and low system pressures and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. The model developed takes into consideration the competing evaporation and condensation effects in the subcooled boiling regime. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended programme is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (flow excursion) in the subcooled boiling regime, comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the programme extension ensures simulation of thermohydraulic instability.