Large Loss Of Coolant Accident Research Articles

Exploratory study of the EU-DEMO Water-Cooled Lithium Lead breeding blanket behaviour in case of loss of cooling capability

Within the framework of the European Roadmap to the realization of fusion energy, a strong international cooperation is ongoing to develop a Breeding Blanket (BB) system for the EU-DEMO reactor. Although it is still to be decided whether the DEMO in-vessel components should perform any safety function, the pursuing of robust blanket concepts able to handle upset and accidental loading conditions has been always seen as good practice in fusion reactor engineering to enhance the inherent plant safety performances. Amongst the several classes of events that might challenge the BB structural integrity, the large Loss of Coolant Accident is one of the most relevant because it usually leads to a fast loss of cooling capability of the structures. Due to the characteristic of the tokamak assembly, the behaviour of each blanket segment during a sudden loss of cooling capability does not depend only upon distinguishing features of the component itself. In fact, the overall transient can be governed by conditions established in surrounding elements, like adjacent blanket segments and vacuum vessel, as well as by the plasma shutdown strategies adopted to protect the reactor.The scope of the activity herein presented is to make a preliminary assessment of the intrinsic capability of EU-DEMO tokamak architecture to cope with the loss of cooling in the Water-Cooled Lithium Lead (WCLL) BB concept. Evaluation of BB thermal field in short and medium term under simplified, yet conservative, assumptions was carried out for four transient scenarios with the aim of investigating the response of the structure in case of: a) fast or soft plasma shutdown, and b) different blanket cooling schemes. Moreover, the WCLL BB thermo-mechanical response in the most critical time steps has been assessed. The obtained results shall help for future decisions on safety systems/action to be implemented to cope with accidents.

Experimental qualification of the ITER pressure suppression system by means of a large scale facility: assessment of similitude analysis

The paper deals with the assessment of the pressure suppression system of the international thermonuclear experimental reactor (ITER). An extensive experimental research programme performed in a small-scale test facility analysed the steam direct condensation at sub-atmospheric pressure (about 400 tests). A similitude analysis was elaborated for extrapolating the results to the ITER full scale system. The derived scale laws were applied to analyse the performances of the pressure suppression system in order to manage a large loss of coolant accident (LOCA) scenario postulated in the vacuum vessel of ITER. Experimental tests, performed in a large-scale test facility built at the University of Pisa (geometrical scale about 1 and 1/10 power scale), analysing the first part of the large LOCA event, assessed the scale laws emphasizing some discrepancies determined in the extrapolation of reduced scale results to the actual system. Introducing a justified correction factor, the presented similitude analysis assured a suitable and reliable temperature and pressure increase prediction of full-scale systems, even in scaled configuration affected by relevant thermal stratification.

Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1

Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.

Improving nuclear power plant safety through independent water storage systems

The safety of nuclear power plants (NPPs) depends on the actions taken to prevent nuclear accidents or to limit their consequences. This paper proposes improvement of pressurized water reactor safety by introducing of new system for water injection in the primary and secondary loops. The system is envisioned as a supplementary to the regular safety systems and to be operated independently. The proposed design consists of a water source, pipes, valves and a portable fossil-fuel driven pump, and implements the “Diverse and Flexible Coping Strategy” mitigation concept recommended by the US industry following the Fukushima accident. The potential of the supplementary water storage system to cope with two major initiating events, i.e. the station blackout and large loss of coolant accident, is investigated, and its broad impact on nuclear safety is quantified. The probabilistic safety assessment is employed in the analyses. In particular, a probabilistic safety assessment model of a pressurized water reactor is modified to consider the new independent water storage system. The core damage frequency of the plant is used as a risk measure. The results reveal that the core damage frequency of the analyzed NPP decreases significantly after the introduction of the proposed water storage system. Furthermore, the criticality of some of the most important basic events related to the provision of water flow during station blackout significantly decreases.

Effects of Filtered Containment Venting on Fission Product Releases During CANDU Reactor Severe Accidents

Severe accidents are of increasing concern in the nuclear industry worldwide since the accidents at Fukushima Daiichi (March 2011). These events have significant consequences that must be mitigated to ensure public and employee safety. Filtered containment venting (FCV) systems are beneficial in this context as they would help to maintain containment integrity while also reducing radionuclide releases to the environment. This paper explores the degree to which filtered containment venting would reduce fission product releases during two Canada Deuterium Uranium (CANDU) 6 severe accident scenarios, namely a station blackout (SBO) and a large loss of coolant accident (LLOCA) (with limited emergency cooling). The effects on the progression of the severe accident and radionuclide releases to the environment are explored using the Modular Accident Analysis Program (MAAP)–CANDU integrated severe accident analysis code. The stylized filtered containment venting system model employed in this study avoids containment failure and significantly reduces radionuclide releases by 95–97% for non-noble gas fission products. Filtered containment venting is shown to be a suitable technology for the mitigation of severe accidents in CANDU, maintaining containment integrity and reducing radionuclide releases to the environment.

Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor

Abstract Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavodâ-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described.

Potential improvement of CANDU NPP safety margins by shortening the response time of shutdown systems using FPGA based implementation

Abstract The relationship between the peak values of critical reactor variables, such as neutronic power, inside a CANDU reactor and the speed of the response of its shutdown system has been analyzed in the event of a large loss of coolant accident (LOCA). The advantage of shortening the response time of the shutdown action has been demonstrated in term of the improved safety margin. A field programmable gate array (FPGA) platform has been chosen to implement such a shutdown system. Hardware-in-the-loop (HIL) simulations have been performed to demonstrate the feasibility of this concept. Furthermore, connections between the speed of response of the shutdown system and the nominal operating power level of the reactor have been drawn to support for potential power upgrade for existing power plants.

Transient analysis of pressure and temperature of fluid into AP1000 reactor containment during DECL accident with several volumes modelling

Since the accidents at Three Mile Island and Chernobyl, more stringent requirements have been imposed on the containment system for full replica nuclear power plant simulators. The simulation must be able to accurately respond under all normal operations and especially under abnormal conditions, including leaks and breaks of any size in any location in the reactor building. This paper presents one of the most dangerous accidents in reactor containments known as Loss of Coolant Accident (LOCA) in its worst condition which is called large LOCA. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means totally guillotine type of break in cold leg pipe. This modelling is performed in the four volume method in the AP1000 reactor, which is one of the most sophisticated safe reactors that have ever been built. The conservation mass, energy and momentum equations have been used in this modelling, and the modelling software applied in our analysis is MATLAB. The results are compared with the AP1000 safety, security and environmental reports.

A Methodology for Determining Interactions in Probabilistic Safety Assessment Models by Varying One Parameter at a Time

In risk analysis problems, the decision-making process is supported by the utilization of quantitative models. Assessing the relevance of interactions is an essential information in the interpretation of model results. By such knowledge, analysts and decisionmakers are able to understand whether risk is apportioned by individual factor contributions or by their joint action. However, models are oftentimes large, requiring a high number of input parameters, and complex, with individual model runs being time consuming. Computational complexity leads analysts to utilize one-parameter-at-a-time sensitivity methods, which prevent one from assessing interactions. In this work, we illustrate a methodology to quantify interactions in probabilistic safety assessment (PSA) models by varying one parameter at a time. The method is based on a property of the functional ANOVA decomposition of a finite change that allows to exactly determine the relevance of factors when considered individually or together with their interactions with all other factors. A set of test cases illustrates the technique. We apply the methodology to the analysis of the core damage frequency of the large loss of coolant accident of a nuclear reactor. Numerical results reveal the nonadditive model structure, allow to quantify the relevance of interactions, and to identify the direction of change (increase or decrease in risk) implied by individual factor variations and by their cooperation.

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4–CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4–CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4–CANDU code predictions.

Modifications in SCDAP code for early phase degradation in a CANDU fuel channel

The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (CANDU) reactor type constitutes the major feature which prevents the direct application to this reactor design of the severe accident physical models developed for PWR/BWRs. During a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency core cooling (LOECC), particular deformation phenomena take place inside a CANDU fuel channel. Another peculiarity appears at melt relocation in horizontal geometry. Brief discussion of accident phenomena is included. SCDAP/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. The paper summarizes the applications to this reactor type previously performed by the authors with the standard (not modified) SCDAP/RELAP5 (RELAP/SCDAPSIM/MOD3.4). To obtain a code version able to simulate a CANDU channel during the early phase of a severe accident, two new models are developed inside the SCDAP routines structure. These models concern the metallic melt relocation inside the horizontal fuel channel and fuel bundle slumping. Major features of the original LIQSOL (Liquefaction-flow-Solidification) model in SCDAP are described, together with the assumptions based on experimental evidence that led to the development of the modified model. Also, two sub-models that form the basis of fuel bundle slumping treatment are presented: (a) in-built resizing of hydrodynamic sub-channels inside the fuel channel, and (b) heat transfer at direct contact, fuel to fuel and fuel to shroud. The effects of the new models on the overall performance of the code are tested in several different cases and against the results of the standard RELAP/SCDAPSIM/MOD3.4(bi7). The plans to validate the new models are discussed, as well as some general adaptation directions to different phases of a severe accident in CANDUs, foreseen for SCDAP.

SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR–BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core. This article presents the analysis of large loss of coolant accident (LOCA) with emergency core cooling system failure and station blackout cases without any operator intervention and when the cooling of reactor control rods is restored. For the station blackout case the other non-regular measures (depressurisation of reactor cooling system, water injection from hydro-accumulators, deaerators and city water system) are evaluated. The analysis was performed using RELAP5 Mod3.2, RELAP/SCDAPSIM and RELAP5-3D codes. The use of three codes allows to investigate all aspects of accident with long-term loss of heat removal from the core: RELAP5 code allows to perform thermal-hydraulic analysis using detailed RBMK-1500 model; RELAP/SCDAPSIM code is necessary for modelling of core damage progression; RELAP5-3D code with special “multidimensional heat conduction” model allows to model heat transfer from hot fuel channels through graphite column in radial direction. Results of the analysis have shown, that the heat removal from control rods cooling circuit allows to slowdown effectively the core heat-up process.

A neutron scatterometer for void-fraction measurement in heated rod-bundle channels under CANDU LOCA conditions

The difficult problem of directly measuring the void fraction of rapidly boiling water in rod-bundle channels, such as those simulating a large loss-of-coolant accident (LOCA) in a nuclear reactor, is overcome by using a fast-neutron scattering device (scatterometer). The neutron scatterometer measures the number of neutrons scattered from a test section exposed to fast neutrons, and relates these measurements to the channel void fraction. The concept of the device and its design features are discussed. Such a scatterometer was constructed and installed on a test section (channel) of the RD-14M Thermalhydraulic Test Facility of Atomic Energy of Canada Limited. Experiments (30-mm diameter break in the inlet header) were conducted to evaluate the scatterometer’s performance under transient conditions. Results of these tests show that the scatterometer can continuously measure the channel void fraction to within ±10% void (95% confidence interval) when corrections for non-linearity and time response are applied.

An application of the new way to prevent core melting in pressure tube reactors (CANDU type)

The pressure tube reactors, especially CANDU type, have a calandria low pressure vessel (near to atmospheric pressure) immersed into a concrete vault filled with water. The accident analysis done by ELFIN-HTCELL code for the channel heat up and by fluid flow PHOENICS code as applied for moderator cooling system efficacy, showed that even the moderator cooling system operates, in some transients sequences where the normal heat sinks are lost, and the top core pressure tubes can reach burst conditions, which means that the fission product secondary retaining barrier gets destroyed, and yet the core can be cooled by water admission through the ruptured tubes from the emergency core cooling system (ECC), if it is available. Otherwise, if in many accident sequences the moderator cooling system remains the ultimate heat sink for the core fuel, and it is not available even from the accident start, a core melt appears. Taking into account the “natural” advantage offered by the presence of both pools in calandria and in the vault, separated by the calandria vessel, the introduction of density locks between them could be a safety passive design solution. When the temperature of moderator water gets higher the density lock cold-hot interface loss stability and thus the density locks get “open” fully permitting the admission of the cool water from the vault pool in calandria. Therefore, by natural circulation the decay heat is transferred via an air-cooling tower, and no mechanical moving parts are needed to open this circuit. Also, if the vault water is borated, it can be used to stop the nuclear reaction when the normal shutdown systems are not available and a positive reactivity coefficient appears, e.g. large loss of coolant accident (LOCA).

Fog Inerting Effects on Hydrogen Combustion in a PWR Ice Condenser Containment

A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H2–air–steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser containment during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H2 by volume).

Boiling Water Reactor Loss-of-Coolant Tests with Two Bundle Loop

Two simulation tests for a boiling water reactor large loss-of-coolant accident (LOCA), conducted in the two bundle loop, were analyzed using the current licensing code system. These tests were recirculation-pump suction-line double-ended break tests. One of these tests assumed failures for LPCS and 1 out of 3 LPC!s, and another test assumed HPCS failure. A main objective of these analyses is to confirm the conservativeness of the licensing analysis models. Conclusions reached from the analyses are as follows : (1) Calculated heater surface temperature begins to rise much earlier than the measured temperature, due to the conservative GEXL model for a LOCA analysis. (2) Calculated heat up rate is higher than the test data, mainly due to neglecting the steam cooling in the analysis. (3) Calculated heater rewetting time is later than the test data, due to neglect of counter current flow limiting at the core inlet, when the measured ECCS flow rate is used in the analysis. It has been confirmed that the current licensing analysis models give a conservative result for peak cladding temperature (PCT), due to the model conservativeness factors pre­ sented above, when the measured data are used in the analysis for the outflow from the system and the inflow to the system.

Boiling water reactor loss-of-coolant tests with two bundle loop. Examination of conservativeness for licensing analysis models using large break test data.

Two simulation tests for a boiling water reactor large loss-of-coolant accident (LOCA), conducted in the two bundle loop, were analyzed using the current licensing code system. These tests were recirculation-pump suction-line double-ended break tests. One of these tests assumed failures for LPCS and 1 out of 3 LPCIs, and another test assumed HPCS failure. A main objective of these analyses is to confirm the conservativeness of the licensing analysis models. Conclusions reached from the analyses are as follows: 1. Calculated heater surface temperature begins to rise much earlier than the measured temperature, due to the conservative GEXL model for a LOCA analysis.2. Calculated heat up rate is higher than the test data, mainly due to neglecting the steam cooling in the analysis.3. Calculated heater rewetting time is later than the test data, due to neglect of counter current flow limiting at the core inlet, when the measured ECCS flow rate is used in the analysis. It has been confirmed that the current licensing analysis models give a conservative result for peak cladding temperature (PCT), due to the model conservativeness factors presented above, when the measured data are used in the analysis for the outflow from the system and the inflow to the system.