High Pressure Coolant Injection Research Articles

A Qualitative Strategy for Fusion of Physics into Empirical Models for Process Anomaly Detection

To facilitate the automated online monitoring of power plants, a systematic and qualitative strategy for anomaly detection is presented. This strategy is essential to provide credible reasoning on why and when an empirical versus hybrid (i.e., physics-supported) approach should be used and to determine the ideal mix of these two approaches for a defined anomaly detection scope. Empirical methods are usually based on pattern, statistical, and causal inference. Hybrid methods include the use of physics models to train and test data methods, reduce data dimensionality, reduce data-model complexity, augment data, and reduce empirical uncertainty; hybrid methods also include the use of data to tune physics models. The presented strategy is driven by key decision points related to data relevance, simple modeling feasibility, data inference, physics-modeling value, data dimensionality, physics knowledge, method of validation, performance, data availability, and suitability for training and testing, cause-effect, entropy inference, and model fitting. The strategy is demonstrated through a pilot use case for the application of anomaly detection to capture a valve packing leak at the high-pressure coolant injection system of a nuclear power plant.

Precautions at Fukushima That Would Have Suppressed the Accident Severity

Abstract The Great East Japan Earthquake on Mar. 11, 2011 triggered huge tsunami waves that attacked Fukushima Daiichi Nuclear Power Plant (Fukushima-1). Units 1, 3, and 4 had hydrogen explosions. Units 1–3 had core meltdowns and released a large amount of radioactive material. Published investigation reports did not explain how the severity of the accident could have been prevented. We formed a study group to find: (A) Was the earthquake-induced huge tsunami predictable at Fukushima-1? (B) If it was predictable, what preparations at Fukushima-1 could have avoided the severity of the accident? Our conclusions were: (a) The tsunami that hit Fukushima-1 was predictable, and (b) the severity could have been avoided if the plant had prepared a set of equipment, and most of all, had exercised actions to take against such tsunami. Necessary preparation included: (1) a number of direct current (DC) batteries, (2) portable underwater pumps, (3) portable alternating current (AC) generators with sufficient gasoline supply, (4) high voltage AC power trucks, and (5) drills against extended loss of all electric power and seawater pumps. This set applied only to this specific accident. A thorough preparation would have added (6) portable compressors, (7) watertight modification to reactor core isolation cooling system (RCIC) and high pressure coolant injection system (HPCI) control and instrumentation, and (8) fire engines for alternate low pressure water injection. Item (5), i.e., to study plans and carry out exercises against the tsunami would have identified all other necessary preparations.

EPRI MAAP5 Fukushima Daiichi Analysis

The study presented in this paper summarizes work conducted as part of the Electric Power Research Institute (EPRI) Fukushima Technical Evaluation project. This effort was designed to develop a representation of the core damage events that occurred at Fukushima Daiichi Units 1, 2, and 3 using the analytical capabilities provided by the EPRI Modular Accident Analysis Program, Version 5 (MAAP5). The analytical investigations of Fukushima Daiichi performed with MAAP5 indicate that core-melt progressions at Units 1, 2, and 3 likely span a range of core damage conditions. The core status at Unit 1 is likely consistent with a large fraction of core debris having relocated into the containment. By contrast, the MAAP5 evaluations indicate that there is a reasonable potential for a significant fraction of core debris to be retained inside the reactor pressure vessel (RPV) at Unit 2. The corresponding Unit 3 simulations, however, highlight the important role that degraded high-pressure coolant injection at low RPV pressure may have played in promoting some relocation of core debris out of the RPV and into containment. The detailed containment evaluations conducted as part of this study also highlight the critical role played by thermal stratification phenomena (either in the suppression pool or in the drywell) in influencing the magnitude of containment pressure and thermal challenges. These simulations highlight the potentially critical role that thermal stratification in the upper drywell may have played in accelerating the onset of leakage through the drywell head flange due to thermal degradation of the drywell head gasket. Finally, these simulations provide good representations of the occurrence of flammable conditions in the Units 1 and 3 reactor buildings, supporting the nature and timing of the observed reactor building combustion events.

Analysis of the accident in the Fukushima Daiichi nuclear power station Unit 3 with MELCOR_2.1

During the major accident occurred at the Fukushima Daiichi nuclear power station in March 2011, three units of the nuclear power plants suffered extensive damage to the reactors and buildings. It is widely believed that all three reactor cores experienced some melting.In the present paper, the Fukushima Unit 3 accident has been studied by using Melcor_2.1. An initial calculation was performed by using design conditions or operators reported actions. Several series of sensitivity cases have been performed in order to reproduce the main accident measurements (e.g. pressure histories of the reactor pressure vessel, dry-well and wet-well; downcomer water levels and the observed hydrogen explosion time). The sensitivity studies consisted in variations of Reactor Core Isolation Cooling and High Pressure Coolant Injection operation, Alternative Water Injection, venting, leaking to the dry-well and failure parameters of the lower head. The outcomes of the sensitivity calculations have pointed out the likely-state of the U3 core after 6days of the accident start.

Phenomena identification and ranking table for thermal-hydraulic phenomena during a small-break LOCA with loss of high pressure injection

Currently Appendix K of 10 CFR 50 is used in the United States to evaluate models for the emergency cooling systems of light-water reactors. To assure that these models are accurate enough to ensure that the cooling systems are satisfactory, code scalability, applicability and uncertainty methodologies are used to evaluate the uncertainty in system analysis code predictions due to the various models. One cornerstone of this methodology is the development of the Phenomenon Identification and Ranking Table (PIRT), which summarizes the thermal-hydraulic phenomenon associate with a particular accident scenario and ranks their importance in determining the effectiveness of core cooling and the parts of the accident for which the phenomenon may be important. In this paper the PIRT developed by the Institute for Nuclear Safety Systems for a small-break LOCA with loss of high-pressure emergency coolant injection is analyzed in detail and several modifications are proposed based on a mechanistic understanding of the phenomenon involved. The resulting PIRT should provide a more accurate guide for model evaluation and development in advanced thermal-hydraulic system analysis codes.

Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3 (Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model)

Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3 (Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model)

Response Spectrum Analysis of the Condensate Storage Tank in a Nuclear Power Plant

Following the nuclear power plant accident in Fukushima Japan, seismic capacity evaluation has become a crucial issue in combination building safety. Condensate storage tanks are designed to supplies water to the condensate transfer pumps, the control rod drive hydraulic system pumps, and the condenser makeup. A separate connection to the condensate storage tank is used to supply water for the high pressure coolant injection system, reactor core isolation cooling system, and core spray system pumps. A condensate storage tank is defined as a seismic class I structure, playing the important role of providing flow to the operational system and the required static head for the suction of the condensate transfer pumps and the normal supply pump. According to the latest nuclear safety requirements, soil structure interaction must be considered in all seismic analyses. This study aims to rebuild the computer model of condensate storage tanks in Taiwan using the SAP 2000 program in conjunction with the lumped mass stick model and to evaluate the soil structure interaction by employing the SASSI 2000 program. The differences between the results with the soil structure interaction and spring model are compared via natural frequency and response spectrum curves. This computer model enables engineers to rapidly evaluate the safety margin of condensate storage tank following the occurrence of earthquakes or tsunamis.

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

Abstract The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics.

On comparing PRA results with operating experience

A procedure is presented for quantifying the consistency between probabilistic risk assessment (PRA) results and corresponding plant-specific operating data not considered in the PRA. The method, which is easily implemented in practice, is based on the use of Bayes p-values for the predictive probability that the observed data would have been produced from the PRA results in conjunction with an assumed binomial or Poisson sampling distribution. Uncertainties in both the PRA results and the operating data are considered. The method is used to quantify the consistency between PRA results and operating data for high-pressure coolant injection system unreliabilities at 11 US commercial boiling water reactors.

Application of Reliability-Centered Maintenance to Boiling Water Reactor Emergency Core Cooling Systems Fault-Tree Analysis

Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system functionbased, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing (as required by Section XI of the American Society of Mechanical Engineers Code) was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., highpressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company’s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992.The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PMperiod failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.A significant decrease in CDF was found when comparing the plant-specific data with the generic data. Only a small difference in CDF was found between plant-specific RCM data compared with PM data, which can be attributed to the few number of essential MOVs modeled in the HPCI, RCIC, RHR/LPCI, and LPCS PRA fault trees.

Influence of Isolation Condensers on Boiling Water Reactor Performance During an Anticipated Transient Without Scram

The purpose of isolation condensers (ICs) in a boiling water reactor (BWR) is to passively control the reactor pressure by removing heat from the system during abnormal reactor operation. This type of control is expected to reduce the frequency of opening and closing of the safety relief valves (SRVs). In addition, the ICs are designed to remove the decay heat of the reactor system after reactor shutdown in case of a loss of coolant accident. Transient calculations of an anticipated transient without scram for a main steam isolation valve (MSIV) closure event were performed using the RAMONA-4B code with prescribed pressure setpoints for the SRVs and given time settings for the MSIV closure. It is demonstrated that the effect of the ICs is to reduce the rate of pressure rise and thereby decrease the cycling frequency of the SRVs. The onset of high-pressure coolant injection is delayed if an IC is functional. An analysis for natural circulation flow in a BWR indicates that the effect of an IC on the transient performance is similar to that for the forced circulation system. In this case, the MSIV closure results in a lower peak pressure because of the lower power level

Assessment of MOV stroke time testing as a condition-directed PM task

This research focuses specifically on motor-operated valves (MOVs) in the Emergency Core Cooling System (ECCS) for a Boiling Water Reactor (BWR). The possible use of stroke time testing as a condition-directed Preventive Maintenance (PM) task for MOVs is evaluated. The MOV failure modes included: packing failures, control switch related failures, mechanical failures, and motor/electrical-related failures. MOV stroke time test data for the MOVs in the High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), Low Pressure Coolant Injection (LPCI), and Low Pressure Core Spray (LPCS) systems were gathered and analyzed from PECO Energy's Peach Bottom Atomic Power Station in order to assess the value of stroke time testing. This analysis shows that MOV stroke time testing should not be used as a condition-directed PM task. Regression analyses and attempts to identify large stroke time deviations failed to predict possible or eminent MOV failures. Stroke time testing should only be considered as ‘go/no-go’ test, and is therefore limited to only proving operability under no-load conditions.

Uncertainty in Poisson event counts and exposure time in rate estimation

In quantifying a plant-specific Poisson event occurrence rate λ in Probabilistic Risk Assessment, it is sometimes the case that either the corresponding Poisson exposure time t or the observed number of events x (or both) are uncertain. We present several methods which account for uncertainties in both x and t when using Bayesian methods to estimate λ . A gamma prior distribution on λ is considered. While the methods formally require numerical integration, a computationally convenient approximation is provided to implement them in practice. A numerical example concerning the rate of failure to operate of the high pressure coolant injection system of commercial boiling water reactors is used to illustrate the methods.

A method for quantifying the reliability of instrumentation for engineered safety features to improve accident sequence recognition in reactor control rooms

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to quantitatively evaluate the reliability of the instrumentation for engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel through the conversion of system fault trees to alarm trees. In the alarm tree, possible states of each instrumented alarm are identified as “true” or “false”. In addition, a “warning” status is also defined and integrated into the alarm analysis routine. The impact of this additional status condition on the Boolean laws used to evaluate the alarm trees is examined. An application is described for a BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents.