Isolation Condenser System Research Articles

Integral simulation of the first 3 weeks of severe accident at Fukushima Daiichi Unit 1 with SOCRAT code. Part I – Qualification of model and simulation of the initial phase

This paper is the first part of the summary overview which presents the results of IBRAE's work over the last 11 years to study the severe accident at Unit 1 of Fukushima Daiichi Nuclear Power Station. The paper is dedicated to the description of the integral model of the Unit 1 with assumptions made, and to its qualification with the available data. The qualification includes the comparison of simulation results with measurements for the initial phase of the accident which starts with the initiating event (earthquake) and ends with tsunami arrival. The important issues discussed are the direct modeling of isolation condenser performance and verification of the possibility to use an integral approach to estimate the nuclide inventory in the core. These questions are of methodological importance for modeling the main phase of the accident.Direct modeling of isolation condenser as part of integral calculation allows verifying the hypothesis about the isolating valves left open inside containment after tsunami arrival and station blackout. The isolation condenser system could influence the development of severe accident by limiting the pressure peak in the reactor at the time of melt relocation into the lower plenum. Thus, verification of the isolation condenser model at the initial stage of the accident justifies the correctness of results obtained for the core degradation phase of the accident.The integral approach in nuclide inventory estimate presumes the need to mix the fuel with different burn-ups in the effective groups of fuel assemblies, which may bring an uncertainty in the nuclide inventory prediction. Quantification of this uncertainty is important for assessment of prediction accuracy for the source term.

A Passive System Technical Specification for CANDU 6 Generation II Reactor

The Fukushima accident led to concerns about enhancement of safety for the new design nuclear plants and also for the existing fleet of nuclear power plants (NPPs) by introducing passive safety systems. An important objective is to increase the grace time for a plant operator to establish an alternate heat sink in the case of a station blackout (SBO) accident. Efforts made by RATEN (https://www.raten.ro/?lang=en) in the frame of H2020 PIACE projects were to implement a passive safety system in the CANadian Deuterium Uranium 6 (CANDU 6) project. In this project, RATEN was in charge of the engineering design and computational modeling aspects, required to integrate a passive safety system in the existing CANDU 6 project. In order to design a passive safety system, a 3-day SBO accident was credited to occur at the Cernavoda Unit 2 NPP, a CANDU 6–type reactor. An isolation condenser (IC) system capable of transporting the total energy produced in the reactor core due to decay heat was designed and modeled. The engineering design solutions were made by RATEN CITON (http://www.citon.ro/english_index.html), and the thermal-hydraulic analysis was performed by RATEN ICN (https://nuclear.ro/en/) using the RELAP5 computer code (https://relap53d.inl.gov/SitePages/Home.aspx) to confirm natural circulation both in the secondary and the primary circuits during the SBO accident and heat transfer capability of the IC with and without noncondensable gases. The passive safety system design consists of four (4 × 33%) closed loop independent circuits, one for each steam generator. Each loop has an IC design to transport 0.66% of nominal thermal power of the reactor. In order to avoid a rapid transient during reactor cooldown, the system is endowed with four noncondensable gas tanks (one for each IC), connected to the outlet header of each IC, provided for reducing the IC heat flux simultaneously with reactor core residual heat decrease. The design concept was adapted to the CANDU 6 reactor power and the specific layout of the Cernavoda site, starting from the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) patent (the demonstrator of lead fast reactor technology) passive system, to increase the plant operator grace time from 23 h (current situation) to more than 72 h.

Reliability assessment of passive systems using artificial neural network based response surface methodology

Many advanced Nuclear Reactor designs deploys passive systems for enhancing safety Passive systems rely on natural driving forces, such as natural circulation, gravity, internal stored energy etc., which do not require external power sources. The driving forces are weak, hence, the phenomenological failures becomes equally important as compared to mechanical failures for passive systems.Substantial efforts are underway towards improving reliability assessment methods for passive systems, however, consensus is not yet reached. Simplified and subjective assumptions made for generating the Response Surface may not necessarily be true for the complex systems.This paper describes ANN Based Response Surface (ANNBRS) methodology developed for reliability assessment of passive systems. The methodology utilizes Principle Component Analysis (PCA) to reduce dimensionality of system and optimizing efforts in designing training data for ANN model. The ANNBRS methodology is applied to Isolation Condenser System (ICS) to demonstrate improvements over the existing Response Surface based approaches.

CFD simulation of boiling flows inside fuel rod bundle of a natural circulation BWR during SBO

Abstract AHWR, designed in India, is a natural circulation based BWR and equipped with many passive features for safety. The reactor utilizes an isolation condenser system for decay heat removal passively in case of SBO. Earlier, it has been established by analysis and experimentally in the scaled test facility that decay heat can be removed for more than 7 days without any safety implications. However, there remains a possibility of hot spots occurring inside the fuel rod cluster of the reactor while removing heat in sub-cooled or single phase operating zones. In this work, CFD simulations are carried out to predict the void and temperature distribution inside the rod bundle to identify the growth of any hot spots during the removal of decay heat. The modeling framework is developed by validating with the results of high pressure flow boiling experiments cited in the literature. The boiling two phase flow is simulated using an Euler-Euler approach with the incorporation of the RPI wall heat flux partitioning model. Performance of various sub models for interfacial momentum and energy transfer has been studied for development of the boiling frame work. The manuscript includes intrinsic details of the CFD model utilized for the simulation of boiling two phase flows inside a complex geometry of rod bundles. The steady state profiles of the complete flow field inside the rod bundle is estimated at two instances of the transient calling as pseudo steady state instances. Predictions of the CFD are also compared with RELAP5 results. RELAP5 under predicts the void fraction at any axial location. Potential sites for hot spots have been observed by the CFD predictions. CFD is able to capture the boiling phenomena occurring inside the complex geometry of rod bundle. The detailed flow field map including the contours of void fraction and temperature at various axial and radial planes shows potential locations for occurrence of hot spot inside the bundle during decay heat removal.

Experimental and computational simulation of thermal stratification in large pools with immersed condenser

This paper deals with experimental and numerical investigations on thermal stratification phenomenon conducted in Integral Test Loop (ITL) scaled test facility for Advanced Heavy Water Reactor (AHWR). Experiments at decay power level were designed to simulate thermal hydraulic performance of Isolation Condenser (IC) system of AHWR at prototypical pressure (70bar) and temperature (285°C). The experiments were simulated using the Best Estimate (BE) thermal hydraulic system code RELAP5. Two different pool side nodalizations were tried for simulation of IC-pool. Nodalization, where the IC-pool was simulated with one lumped volume did not simulate the observed thermal stratification behavior appropriately. However, it simulated the Main Heat Transport System (MHTS) behavior well. To simulate pool side stratification a three volume pool nodalization was adopted, which simulated the observed thermal stratification behavior in the pool well apart from simulating MHTS system behavior. Study shows a characteristic thermal stratification behavior in pool, which prevails for the initial few hours of IC functioning. The characteristic curve showing various stages of observed pool thermal stratification development and decay during experiment has been presented in the paper. An interrelation between pressure and water level into the pool has been observed in studies.

Development of the APSRA+ Methodology for Passive System Reliability Analysis and Its Application to the Passive Isolation Condenser System of an Advanced Reactor

Research in the field of passive system reliability analysis is garnering sharp interest in the nuclear community. Passive systems are being utilized extensively in current- and future-generation reactors for their normal operations as well as for safety critical operations during any accidental conditions. In this paper, we present a methodology called Analysis of Passive System ReliAbility Plus (APSRA+) for evaluating reliability of passive systems. This methodology is an improved version of the existing APSRA methodology. The methodology has been applied to the passive isolation condenser system (ICS) of the AHWR (Advanced Heavy Water Reactor). With the help of the APSRA+ methodology, the probability of the passive ICS failing to maintain the clad temperature under 400°C is estimated to be of the order 1×10−10.Important features of APSRA+ are the following. First, it provides an integrated dynamic reliability method for the consistent treatment of dynamic failure characteristics such as multistate failure, fault increment, and time-dependent failure rate of components of passive systems. Second, this methodology overcomes the issue of process parameter treatment by just the probability density function or by root cause analysis, by segregating the parameters into dependent and independent process parameters and then giving a proper treatment to each of them separately. Third, the methodology treats the model uncertainties and independent process parameter variations in a consistent manner.In APSRA+, the important parameters affecting the passive system under consideration are identified using sensitivity analysis. To evaluate the system performance, a best-estimate system code is used with due consideration of the uncertainties in empirical models. A failure surface is generated by varying all the identified important parameters; variation from the nominal values of these parameters affects the system performance significantly. These parameters are then segregated into dependent and independent categories. For dependent parameters, it is attributed that the variations of process parameters are mainly due to malfunction of mechanical components or control systems, and hence, root cause analysis is performed. The probability of these dependent parameter variations is estimated using a dynamic reliability methodology based on Monte Carlo simulation. The dynamic failure characteristics of the identified causal component/system are accounted for in calculating these probabilities. For the treatment of independent process parameters, using APSRA+ suggests adopting and integrating classical data-fitting techniques or mathematical models. In the next steps, a response surface-based metamodel is formulated using the generated failure points. The probability of the system being in the failure zone is estimated by sampling and analyzing a sufficiently large number of samples for all the dependent and independent process parameters based on the probability of variations of these parameters, which were estimated using dynamic reliability methodology.

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.

Assessment of passive safety system performance under main steam line break accident

A generation III+Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features which require no emergency injection pump and no operator action or Alternating Current (AC) power supply. The generation III+BWR’s passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS), and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A Main Steam Line Break (MSLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+BWR. The main results of PUMA MSLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the minimum water level (1.706m) was 5% higher than the TAF (1.623m) and the containment maximum pressure (271kPa) was 35% lower than the safety limit (414kPa), respectively.

Reliability assessment of passive isolation condenser system of AHWR using APSRA methodology

In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) is used for evaluation of reliability of passive isolation condenser system of the Indian Advanced Heavy Water Reactor (AHWR). As per the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to perform the design basis function. The methodology first determines the operational characteristics of the system and the failure conditions based on a predetermined failure criterion. The parameters that could degrade the system performance are identified and considered for analysis. Different modes of failure and their cause are identified. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the isolation condenser system performance. Once the failure surface of the system is predicted, the causes of failure are examined through root diagnosis, which occur mainly due to failure of mechanical components. Reliability of the system is evaluated through a classical PSA treatment based on the failure probability of the components using generic data.

Design, experiments and Relap5 code calculations for the perseo facility

Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories. The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe. ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained. An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses. This paper deals with the experimental and calculated results limited to the Relap5 code.

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200. The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations. Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.

Steam Line Break and Station Blackout Transients for Proliferation-Resistant Hexagonal Tight Lattice Boiling Water Reactor

Safety analyses of a proliferation-resistant, economically competitive, high-conversion boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems, are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core that therefore operates with a fast reactor neutron spectrum and a considerably improved neutron economy compared to the current generation of light water reactors. The tight lattice core has a very narrow flow channel with a hydraulic diameter less than half of the regular boiling water reactor (BWR) core and, thus, presents a special challenge to core cooling because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator-to-fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios.In the preliminary HCBWR design, the core is placed in a vessel with a large chimney section, and the vessel is connected to the isolation condenser system (ICS). The vessel is placed in containment with the gravity driven cooling system (GDCS) and passive containment cooling system (PCCS) in a configuration similar to General Electric’s simplified BWR (SBWR). The safety systems are similar to those of the SBWR; the ICS and PCCS are scaled with power. An internal recirculation pump is placed in the downcomer to augment the buoyancy head provided by the chimney since the buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel as the tight lattice configuration results in much larger friction in the core than with the SBWR.The constitutive relationships for RELAP5 are assessed for narrow channels, and as a result the heat transfer package is modified. The modified RELAP5 is used to simulate and analyze two of the most limiting events for a tight pitch lattice core: the station blackout and the main-steam-line-break events. The results of the analyses indicate that the HCBWR system will be safely brought to the shutdown condition for these transients.

Reliability evaluation of a natural circulation system

This paper discusses a reliability study performed with reference to a passive thermohydraulic natural circulation (NC) system, named TTL-1. A methodology based on probabilistic techniques has been applied with the main purpose to optimize the system design. The obtained results have been adopted to estimate the thermal-hydraulic reliability (TH-R) of the same system. A total of 29 relevant parameters (including nominal values and plausible ranges of variations) affecting the design and the NC performance of the TTL-1 loop are identified and a probability of occurrence is assigned for each value based on expert judgment. Following procedures established for the uncertainty evaluation of thermal-hydraulic system codes results, 137 system configurations have been selected and each configuration has been analyzed via the Relap5 best-estimate code. The reference system configuration and the failure criteria derived from the “mission” of the passive system are adopted for the evaluation of the system TH-R. Four different definitions of a less-than-unity “reliability-values” (where unity represents the maximum achievable reliability) are proposed for the performance of the selected passive system. This is normally considered fully reliable, i.e. reliability-value equal one, in typical Probabilistic Safety Assessment (PSA) applications in nuclear reactor safety. The two ‘point’ TH-R values for the considered NC system were found equal to 0.70 and 0.85, i.e. values comparable with the reliability of a pump installed in an “equivalent” forced circulation (active) system having the same “mission.” The design optimization study was completed by a regression analysis addressing the output of the 137 calculations: heat losses, undetected leakage, loop length, riser diameter, and equivalent diameter of the test section have been found as the most important parameters bringing to the optimal system design and affecting the TH-R. As added values for this work, the comparison has been made between results from this study and results from a previous analysis where the same methodology was adopted for the evaluation of the TH-R of a different passive system named Isolation Condenser (IC). The comparison shows that the current single-phase NC system is ‘more reliable’ than the two-phase IC system. This constitutes a proof of qualification and of consistency for the adopted methodology.

Passive System Reliability Analysis: A Study on the Isolation Condenser

This paper deals with the reliability assessment of passive systems that have been developed in recent years by suppliers, industries, utilities, and research organizations, aimed at plant safety improvement and substantial simplification in its implementation. The present study concerns the passive decay heat removal systems that use, for the most part, a condenser immersed in a cooling pool. The focus of the paper is a reliability study of the isolation condenser system foreseen for advanced boiling water reactors (BWRs) for the removal of the excess sensible and core decay heat from the BWR by natural circulation. Furthermore, an approach aimed at the thermal-hydraulic performance assessment (i.e., the natural circulation failure evaluation) from the probability point of view is given. The study is not plant-specific-related but pertains to the conceptual design of the foregoing system.