Loss Of Flow Accident Research Articles

Preliminary safety evaluation of a supercritical carbon dioxide Brayton cycle cooled reactor system under loss-of-flow accident

Reactor system which applies supercritical carbon dioxide(s-CO2) Brayton Cycle as its direct power conversion system has advantages of high thermal efficiency and high compactness. Due to the high coolant outlet temperature (around 500 °C) and the less effective heat transfer capacity of S-CO2 compared to water or metal liquid, which are commonly used in GEN IV reactor concepts, the safety performance of this kind of newly developed reactor system needs to be carefully evaluated. Preliminary safety assessment of loss-of-flow accident (LOFA) for a s-CO2 Brayton cycle cooled reactor system is carried out in this paper. The safety analysis methodology is proposed firstly and three types of LOFA are investigated to clarify the unique safety characteristics of s-CO2 Brayton cycle cooled reactor system, including partial LOFA with reactor scram, partial LOFA without reactor scram and total LOFA. The short term analysis results demonstrate that the reactor core can survive these three types of LOFA without safety system in the very first 10 to 20 s. A hybrid passive/active emergency core cooling system is designed to remove the decay heat for a long time. Results indicate that the hybrid passive/active emergency core cooling system designed in this paper can help the reactor system safely go through these three types of LOFA. This study will be benefit for the core design and safety system design for s-CO2 direct cooled rector system.

Safety analysis of the DONES primary heat removal system

Abstract The development of a neutron source able to reproduce the irradiation conditions typical of a nuclear fusion reactor, in order to test candidate structural materials, is the main goal of the Work Package Early Neutron Source (WPENS) of the EUROfusion action. This source, named Demo Oriented NEutron Source (DONES), is a facility where neutrons are produced by means of D-Li interactions. More in detail, a beam of 125 mA deuterium ions at the energy of 40 MeV strikes a lithium jet flowing in a purposely shaped channel in order to obtain an intense and stable neutron flux for the irradiation of material samples. In the framework of these activities, safety analyses are a key aspect in the DONES design and development. Among the postulated initiating events identified during the preliminary Failure Mode Analysis, the Loss of Flow Accident (LOFA) in the Primary Heat Removal System of the lithium loop, due to a trip of the electro-magnetic pump, is one of the most severe. In fact, the loss of lithium flow, combined with the failed stop of the accelerator, could lead to the destruction of the lithium flow channel in correspondence of the component named Back-Plate. For this reason, it has been chosen to investigate the LOFA adopting the deterministic system code RELAP5-3D. Results, obtained from this analysis, are critically discussed and compared with those obtained by a similar calculation carried out with MELCOR 1.8.6 code, in order to assess the RELAP5-3D capability of describing systems adopting lithium as working fluid.

Integrated deterministic and probabilistic safety assessment of a superconducting magnet cryogenic cooling circuit for nuclear fusion applications

The most promising configuration of a nuclear energy fusion system is the tokamak, the largest of which, called ITER, is under construction in Cadarache, France, which uses a complex system of superconducting magnets to generate a field of several tesla (T), aimed at confining the plasma in the toroidal chamber where nuclear fusion reactions occur. For industrial development, the safety of nuclear fusion systems has to be proved and verified by a systematic analysis of operational transients and accidental conditions. Although the final aim of fusion reactors is to reach steady state operation, present-day tokamaks present complex dynamic features, as their operation is based on the transformer principle with a subset of the superconducting magnets operating in a pulsed mode, to inductively generate plasma currents of the order of several MA. We adopt the framework of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA), for identifying the component failures that may cause a Loss-Of-Flow-Accident (LOFA) in the cooling circuit of a superconducting magnet for fusion applications. Post-processing of the simulated scenarios for the identification of the abnormal transients is performed in an unsupervised manner resorting to a spectral clustering approach embedding a Fuzzy-C Means (FCM) that is compared with an Extended Symbolic Aggregate approximation (ESAX) from the literature that also resorts to the FCM for the classification.The proposed approach turns out to be more efficient than ESAX in the identification of clusters and “prototypical states” of abnormal system behavior. Results show that none of the identified scenarios (even those leading to a LOFA) are critical for the ITER central solenoid module integrity, in the mode of operation considered.

Application of the best-estimate model calibration and prediction through experimental data assimilation methodology to the tests performed on a helium cooled First Wall Mock-up

Abstract Experiments are under progress at the Karlsruhe Institute of Technology (KIT) as part of the general strategy for the development of the DEMO Helium Cooled Pebble Bed (HCPB) design. These experiments are part of the program on the qualification of components and on the validation of system codes. In 2018, thermal-hydraulic investigations have been performed looking into the transient response of a helium-cooled blanket First Wall Mock-Up under Loss Of Flow Accident (LOFA) conditions. The results obtained through these investigations captured the quick temperature rise on the surfaces exposed to heat loads (such as those coming from the plasma), and provided a robust set of data for the validation of system codes such as RELAP5-3D. Then, the present work describes the validation activities performed on the RELAP5-3D code. This validation was performed employing also a Best-Estimate methodology called “Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology”. In a previous work, the same methodology was used to calibrate a RELAP5-3D model of the HELOKA-HP electrical heater under normal operation. In this regard, the present work marks the first application of this methodology to fast transients in the field of fusion reactors.

Test Facility and Results of ITER PF4 Current Leads

The International Thermonuclear Experimental Reactor (ITER) high-temperature superconducting (HTS) current leads designed and manufactured by Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP), Hefei, China, are used to transmit current to the magnetic coil of the ITER device with lower heat conduction load. PF4 HTS current leads which supply pulse current for PF4 coil are the first pair of all 33 pairs series current leads, have finished its test recently. The cold test was performed in a new test facility in which the current leads were vertically installed, and a dedicated design was added to abate the gravitation produced between the two current leads. Besides the cold test facility, this article will also summarize the major test results, joint resistance, high current-sharing temperature, loss of flow accident (LOFA) time, overheating time, heat conduction loads to 5-K ends, and the consumption of 50-K mass flow.

Safety analysis of the 100 kA HTS current lead for CFETR

Based on the Agreement of Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Ministry of Science and Technology of China during the Thirteen Five-Years Plan (so called 13∙5 plan), a dedicated test facility of superconducting magnet for China Fusion Engineering Test Reactor (CFETR) will be constructed in Hefei, China. A pair of high temperature superconducting (HTS) current leads rated at 100 kA for the CFETR magnet test facility is being designed at ASIPP. Because magnetic energy stored in the toroidal field magnet is 134 Giga Joules which are 3 times higher than the ITER case, HTS current leads must work at nominal current without performance degradation during the HTS current lead quench. In order to verify the safety and reliability of HTS current lead, the analysis of static magnetic field is done by ANSYS iteration to predict the maximum current carrying capacity of HTS module. This paper also give the numerical calculation of loss of flow accident(LOFA) of HTS current lead and temperature margin analysis of HTS module. The analysis results indicate that the HTS module has a current carrying capacity of 165 kA without external field at 65 K and a current carrying capacity of 149 kA with external 30 mT vertical field at 65 K. The LOFA time is about 580 s and the burnout time is about 30 s, which is better than the requirement of the toroidal field coil fast discharge time of 24 s.

CIRCE-ICE PLOHS experimental campaign

In the framework of the Heavy Liquid Metal (HLM) - GEN IV Nuclear reactor development one of the main tasks deals with the thermalhydraulics analysis of Heavy Liquid Metal (HLM) cooled fast reactors (attention mainly focused on LBE and Lead-cooled reactors), aimed at their safety analysis in response to a hypothetical accidental scenario. To support this task, the experimental campaign on the CIRCE-ICE configuration (funded by the EU project SESAME WP3) was developed and accomplished at the ENEA Brasimone Nuclear laboratory. The dataset produced represents a database for thermal hydraulic design of components for future HLM Fast Reactors. The CIRCulation Eutectic (CIRCE) is a pool type experimental facility working with Lead Bismuth Eutectic (LBE) as coolant (total inventory up to 90 tons). Inside the CIRCE vessel, the test section named Integral Circulation Experiments (ICE) was hosted and operated in order to investigate thermal hydraulic phenomena occurring during abnormal conditions (Protected Loss Of Heat Sink with Loss Of Flow Accident (PLOHS + LOFA)). This document describes the achieved experimental results of the proposed experimental campaign for the full power steady state condition (normal operation conditions) and for the transient phase (transition to decay heat removal conditions). Experimental data relating to thermal hydraulic phenomena such as the modification of the thermal stratification in “pool type” configuration, coolant mass flow rate modification in the test section occurring during the simulation of the designed accidental conditions and the capability to cool the 37-pins fuel bundle under natural circulation conditions are here reported and described.

Loss of flow accident and loss of heat sink accident analyses of the WCCB primary heat transfer system for CFETR

The Primary Heat Transfer System (PHTS) of the Water Cooled Ceramic Breeder (WCCB) blanket was designed to fulfill multiple operation modes of Chinese Fusion Engineering Testing Reactor (CFETR) with the fusion power of 200 MW in Phase I and 1.5 GW in Phase II. One PHTS loop was modeled using RELAP5. Loss of Flow Accident (LOFA) and Loss of Heat Sink (LOHS) were analyzed to evaluate the reliability of the PHTS design. Both accidents were simulated under two phases of CFETR with two postulations. The first postulation was that accidents happened without detection. Results showed that the blanket was overheated. The second postulation was to verify the effectiveness of corresponding counter-measures. For LOFA, if the plasma shuts down in time, the system can establish natural circulation and remove the decay heat effectively. For LOHS, plasma shutdown in time and feed water restoration at appropriate time are effective mitigation measures.

Investigation on RELAP5-3D© capability to predict thermal stratification in liquid metal pool-type system and comparison with experimental data

A numerical activity, aimed to evaluate the capability of RELAP5-3D© to reproduce the main thermal-hydraulic phenomena in an HLM pool-type facility, in different operative conditions, is presented. For this purpose, the experimental campaign performed in CIRCE-ICE test facility has been selected for the code assessment. Two experimental tests have been analyzed: TEST A consisting in a transition from no-power to a full power steady state conditions, and TEST I, consisting in a transition from gas-enhanced circulation to natural circulation, simulating a protected loss of heat sink plus a loss of flow accident. Three different pool modelling approaches are presented, consisting in a single vertical pipe, parallel pipes with cross junctions and multi-dimensional component. The comparison with experimental data has highlighted the need to divide the large pool in several sections to reproduce the natural convection, strictly correlated with the thermal stratification. The multi-dimensional component seems to be the best practice for the evaluation of this phenomenon even if the lack of specific correlation for heat transfer coefficient in quasi-stagnant conditions in large tanks is a limit for the accuracy of the results. In addition, the paper presents a detailed nodalization of the fuel pin bundle, highlighting quite good capabilities of RELAP5-3D as a subchannel analysis code.

Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Test Reactor (CFETR). The core parameters of CFETR have been updated to major radius R = 7.2 m, minor radius a = 2.2 in 2018. Accordingly, the WCCB blanket design was also updated to satisfy multiple operation modes of CFETR with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. The WCCB blanket is characterized by 3 independent cooling systems (CSs), namely CS1 for the cooling of the First Wall (FW), CS2&3 for Breeding Zones (BZs). At 200 MW of low fusion power, CS3 is idle. At 500 MW to 1.5 GW of high fusion power, all 3 CSs are put into use to ensure adequate cooling. So it is indispensable to evaluate the feasibility of the updated WCCB blanket from the thermal hydraulic and safety point of view. A system analysis code RELAP5, was employed to model the blanket. The steady states were analyzed under 200 MW, 500 MW, 1.0 GW and 1.5 GW of fusion power. Then, the Loss of Flow Accident (LOFA) was simulated with different assumptions to verify the robustness of 3 cooling systems design under different initial postulations. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature, can basically meet the design criterion, which preliminarily verifies the feasibility of the WCCB blanket from the thermal hydraulic safety point of view. Comprehensive system design and analysis is under way.

Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code

Abstract In this paper, a 5 MW open pool type research reactor is simulated and analyzed to check and pass a complete and practical safety assessment against various LOFA scenarios. Generally, LOFA may occur in a research reactor when the reactor primary pump or the safety flapper valve fail. The simulation includes the study of a downward core cooling system circulated by the gravity driven force, the performance and failure effects of the safety flapper valve, the flow reversal mode and transition, the natural convection mode, different possible types of LOFA, and impacts of the surrounding empty boxes. First of all, the code nodalization is successfully benchmarked against available experimental data of the reactor operating parameters. Then, all possible DBAs in this field of study are simulated and discussed in detail. Results are completely satisfactory to simulate and analyze a pool-type research reactor in response to LOFA using the RELAP5 code. Furthermore, transients including the natural convection mode and even flow reversal mode are following naturally without any oscillation or source of errors. Finally, TRR is completely safe against the DBA type of LOFA scenarios.

Scaling analysis on LOFA for the test model of WCCB blanket in CFETR

The water Cooled Ceramic Breeder (WCCB) blanket, as a candidate for Chinese Fusion Engineering Test Reactor (CFETR), must demonstrate its safety before engineering application. Enough data are necessary and the test model should ensure that the data are convincing within reasonable expenses. In this paper, scaling criteria are identified for loss of flow accident (LOFA) in WCCB blanket coolant system, which guides the design scheme of test model that the length scaling factor of 1 and volume scaling factor of 1/120. Scaling analyses are performed to develop similarity criteria for pump coastdown and loop natural circulation in three temporal phases identified by LOFA simulation. The test model only preserves a branch at the sector level and sub-module level, which results in different volume scaling factors at levels. Typical heat-transferring cells, constituted by blanket components, are preserved for heat transfer similarity. By computational simulation of an ideal model, it is testified that the test model can reproduce thermal-hydraulic responses within acceptable range of the distortion.

Transient thermal-hydraulic analysis of thermionic space reactor TOPAZ-II with modified RELAP5

Nuclear power and thermionic conversion can serve as a compact, durable energy source for the space exploration and exploitation. In this paper, the modified Reactor Excursion and Leak Analysis Program5 (RELAP5) with the implement of NaK-78 eutectic alloy (78%K and 22%Na) properties and heat transfer correlations is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including the thermionic fuel elements (TFEs), reactor core, radiator, coolant loop and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, reflector, moderator and the reactivity insertion effects of control drums and safety drums are considered. The steady state condition and three severe transient accidents including reactivity insertion accident (RIA), loss of flow accident (LOFA) and loss of coolant accident (LOCA), are simulated and analyzed. The steady state calculated results agree well with the design values. During the three accidents, the moderator plays a dominant role in the positive temperature reactivity feedback. The coolant has at least 50 K temperature margin to the boiling point. The fuel and TFE components are all below their melting temperature. The progress of these accidents provide relatively sufficient time for operator's response. The calculation results prove that the reactor is a safe and reliable system.

Probabilistic evaluation of the energetics upper bound during the transition phase of an unprotected loss of flow accident for a sodium cooled fast reactor by using a Phenomenological Relationship Diagram

Abstract One of the main research goals of the GEN-IV systems is enhancing their safety compared with the former Sodium-Cooled Fast Reactor (SFR) designs. A key issue is the capability of accidents prevention as well as of demonstrating that their consequences do not violate the safety criteria. In order to fulfill such requirements, risk analyses of severe core disruptive accidents are performed. Since the beginning of the SFR development, Hypothetical Core Disruptive Accidents (HCDAs) have played an outstanding role. Numerous safety analyses have been performed for developing and licensing past SFR designs and nowadays a large database of results is available. In particular, a large amount of results of the mechanistic SIMMER-II and SIMMER-III/IV analyses for various core designs and different power classes is available at the Karlsruhe Institute of Technology (KIT). The current paper describes the probabilistic approach based on the Phenomenological Relationship Diagram (PRD), which is used to evaluate the Probability Distribution Function (PDF) of the thermal energy release during the transition phase of an unprotected loss of flow accident scenario for a SFR. The technique allows taking into account the mechanistic nature of the accident scenario. In fact, the available results of the mechanistic analyses of HCDAs in SFRs are used to assess the PDFs of the dominant phenomena affecting the thermal energy release, which are propagated in the PRD by employing a Monte Carlo method.

Temperature and velocity instrumentation and measurements within a separate-effects facility representing modular reactor core

Temperature and velocity sophisticated measurement techniques have been integrated in a novel way for plenum-to-plenum natural circulation thermal hydraulics investigations in a separate-effects dual-channel facility designed with a representative geometry of prismatic modular reactor core. Time-and-frequency domain analysis has been conducted for the time series data obtained to determine the appropriate sampling time and frequency needed to collect data that statistically represents the time-averaged mean value. Preliminary tests have been conducted to examine the validity of implementing these techniques as well to investigate if flow fields inside prismatic core heated channels are affected by other different temperatures core channels. It is worth mentioning that the velocity measurements (i.e., obtained by using hot wire anemometry) were corrected to account for temperature differences between calibration and experimental conditions. Preliminary results reveal that the top section of the heated channel is significantly affected by cooled channel. This implies that naturally-driven flow has a delicate nature and flow fields inside core heated channels are influenced by temperature gradients exist within the prismatic reactor core after loss of flow accidents.

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria). The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor. The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior. 3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.

A two-phase model for simulation of MTR type research reactor during protected and unprotected LOFA

An optimized, accurate and fast model is investigated to simulate the transient behavior of a research reactor during fast and slow LOFA (loss of flow accident). Simulation is performed in two states when the reactor scram system is initiated at critical core power and when it doesn't work properly. The latter case leads to boiling condition and hence the reactor core will be encountered with two-phase flow condition. A lumped heat conduction model for fuel and clad accompanied with a two-phase flow model for coolant are coupled with point kinetic equations to simulate reactor behavior during the accident. A combination of Runge-Kutta method and up-wind finite difference technique are adopted for the solution to the set of differential equations. The calculation performed for IAEA 10 MW benchmark problem in HEU and LEU fuel. Some important parameters such as the peak temperatures of fuel and clad are obtained for both HEU and LEU conditions during transients. Results of numerical simulations, at the first, agree well with the general physical rules and also verified through comparison with the other computer code results. According to present simulation, under both protected and unprotected LOFA transient, the peak clad surface temperatures remain in safe mode.

Reactivity feedback effect on loss of flow accident in PWR

Abstract In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.

Quantification of initiating event frequencies and component reliability data in level 1 probabilistic safety assessment at Puspati TRIGA research reactor

Prior to developing level-I probabilistic safety assessment in Puspati TRIGA Reactor (RTP), collecting and analyzing data is the cornerstone of the probabilistic safety assessment processes. This paper provides on how data quantification is made available for both initial initiating events frequency and components reliability. Taking into consideration on the availability of data, it was decided that generic data to be used in combination with plant-specific data. Methods for estimating the parameters used in quantifying is also presented in this paper. This includes, data treatment by applying Bayesian updates in forming a better refine posterior data. It was found out that, result with the presence of plant-specific data history tends to have high value, loss of offsite power (LOOP) with 1.27E-03. Meanwhile, for other IEs that used generic data, shown lower result with maximum value of 2.68E-06 (loss of flow accident at secondary pump: LOSC-P). As for components data, basic event, No signal from fission chamber 1 (fail to function) is the highest value with 4.07E-01, followed with Manual valve SV1A fail to close (2.75E-01). This publication is perhaps one of the first publically available document providing information on generic and plant specific data for initiating event and components reliability parameters that were used in a Level-1 PSA study for a nuclear research reactor. In addition, this information can be used by PSA technical staff in how to analyze and treat initial incident event data, including, type and what data need to be collected.

Development and validation of the blanket First Wall mock-up model in RELAP5-3D

In the frame of the development of the equatorial outboard blanket module of the EU DEMO (Demonstration Fusion Power Plant) blanket concepts, the qualification and testing of mock-ups are foreseen by means of different experiments to validate models and computer codes for the simulation of abnormal and accidental operations. In this context, a functional test mock-up of the first wall (FWMU) is planned to be integrated in the HELOKA facility (HElium LOop KArlsruhe) at KIT in order to investigate typical safety relevant transients such as a LOFA (Loss Of Flow Accident). This work is aimed at the development and validation of a detailed thermal-hydraulic model of the FWMU plate by means of the system code RELAP5-3D. To this purpose, a comparison with CFX analyses for both steady-state and transient conditions has been performed as well as a calibration with the pressure loss characteristics experimentally determined under air flow conditions. The present model will be integrated into the completed HELOKA loop model to simulate different operational scenarios, but also to support the definition of a suitable text matrix, set-up the control and to define the instrumentation needed for the experiment.