Accident Management Procedures Research Articles

Implementation of the design extension conditions concept in the Czech Republic regulatory framework and NPPs

The paper describes evolution of the concept of design extension condition (DEC) and its implementation in the Czech Republic. The concept of design extension condition had been introduced in the early 1990ies and later was accelerated after the Fukushima accident in 2011. Today, the concept is elaborated in the EUR document, in the WENRA Reference Levels, and also in the IAEA Safety Standards. A number of countries are implementing the DEC concept into the national regulatory framework, safety analyses reports, plant design modifications, and the accident management procedures. Typically, the DEC events are divided into two groups: the DEC without severe fuel damage (DEC-A) and the DEC with severe fuel damage (DEC-B). Information on the implementation of the DEC concept in the Czech Republic, an overview of relevant plant modifications, and examples of safety assessment are given in the paper.

Iodine source term assessment as result of iodine spiking and mass transfer phenomena during a SGTR transient using MELCOR 2.2 and CATHARE 2 codes

The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies.The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools.The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MWe PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters.The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE.The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail.

Cognitive Behavioral Model of an Operation Crew in the Main Control Room of a Nuclear Power Plant Based on a State-Oriented Procedure

The team’s cognitive behavior plays a crucial role in dealing with accidents at nuclear power plants. Herein, the main behaviors of reactor operators and coordinators in performing accident management were analyzed in executing a state-oriented procedure. According to these cognitive behavioral characteristics, we established cognitive behavioral models of accident management procedures. After that, a cognitive behavioral model was established for the team in the main control room of the nuclear power plant based on the two models, which is expected to provide support to the optimization of a corresponding Human Reliability Analysis model.

Severe Accident Management Systems and Procedures

A nuclear power plant’s safe operation involves the planning for non-standard operational emergencies, where pre-determined safety measures and damage control interventions must be taken into consideration depending on the developed event. A nuclear power plant’s safe operation involves the planning for non-standard operational emergencies, where pre-determined safety measures and damage control interventions must be taken into consideration depending on the developed event. The definition of the severe accident management cannot be explained in a single concept, it needs to be examined in detail. As a result of this it becomes necessary to specify the procedure guidelines appropriate to the nature of the event, which can be used with optimal efficiency under hierarchical organisational control. The experience of nuclear accidents in the world and the precognition of future events, the knowledge of existing guidelines for severe accident management needs to be deepened continuously, but at the same time it can be parallelly upgraded with the application of new technologies.

Parameter study at the PKL test facility on heat transfer mechanisms in the steam generator in presence of nitrogen, steam and liquid

PKL is a worldwide unique integral test facility for simulation of thermal-hydraulic phenomena, which can occur in Pressurized Water Reactor under abnormal or accident conditions. The test rig replicates the entire primary side and the relevant parts of the secondary side of western-type PWR plant in the scale of 1:1 in heights and 1:145 in volume and power. The analyses that have been carried out for the last 40 years have always reflected safety issues addressed in the debate of nuclear community. Among goals of experimental programs conducted at the facility there are demonstration of safety margins, investigation of thermal-hydraulic behavior of a plant under accident conditions, verification of accident management procedures as well as provision of data for thermal-hydraulic codes validation. This paper provides a detailed analysis of experiment dedicated to investigation of heat transfer mechanisms in steam generator in presence of non-condensable gas, steam and liquid. Such conditions can occur in a PWR in case of loss of residual heat removal system under cold shutdown conditions. Probabilistic safety assessments indicated that this accident contributes to much larger core damage frequency than originally anticipated. Therefore, an increasing attention has been paid to this accident scenario in the recent years.

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 [1]. By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR) and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN (China General Nuclear) developed code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

Research on Improvement of FQNPP Reactor Protection System Hardware Logic Design

Fuqing Nuclear Power Plant Reactor Protection System has designed emergency control panel ECP, which is used to deal with the risk of loss of digital automatic protection functions due to common software failure. ECP is designed and implemented by analogy technology, which can bypass the computerized DCS system and use buttons, indicator lights, relays and other devices to achieve system-level protection functions. In practical engineering applications, it is found that the current ECP can not meet the requirements of post-accident control in some specific circumstances. Through the analysis of system design and accident management procedures, two design improvements are listed, compared and analysed, and the optimal improvement is given in this paper.

Discussion on the accident behavior and accident management of the HTGR

High Temperature Gas-cooled Reactor (HTGR), which plays an important role in the worldwide development of Generation-IV nuclear energy technology, is well-known for the high safety, high efficiency and process heat application. A commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) has been designed and is now under construction in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in the end of 2020.In this paper, the reactor behaviors of the HTR-PM during the typical design basis accidents (DBAs) and beyond design basis accidents (BDBAs) have been introduced. In the DBAs, the maximum fuel temperature will never exceed its design limitation of 1620 °C, below which under the design burn-up the “Tristructural-isotropic” (TRISO) coated particles can effectively retain their fission-product inventory. Even in the typical BDBAs with extremely low probability, there is enough time, e.g. several days, to adopt appropriate measures to mitigate the consequence, so that the large release of the radioactive materials would not happen.Accident management is important for the nuclear power plant. Besides, After the Fukushima Dai-ichi nuclear accident, in the worldwide, people are more concerned about the severe accident management of the nuclear power plant, and some standards and guidelines are issued. Based on the accident analyses of the HTR-PM, its accident management is studied and discussed. The designers believe that the accident management procedures can be simplified and no offsite emergency measures are needed for the HTR-PM due to the inherent safety design. Compared to the other types of the nuclear power plant, e.g. the Pressurized-Water Reactor (PWR) power plant, the Severe Accident Management Guideline (SAMG) can be simplified or even unnecessary for the HTR-PM.Above work also can provide reference for the further study on the safety regulations, standards or guidelines for the design, operation and accident management of the HTGR.

Three dimensional void distribution measurement of salt-water pool-boiling in 5 × 5 bundle geometry with X-ray CT system

Boiling of sea water may occur in a pressure vessel of light water (nuclear) reactors to flood the nuclear fuel as an accident management procedure. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Boiling behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with three different fluids: water, condensed (two and half times higher concentration) sea water and its mixture solution of sea water and borated water. Three-dimensional void-fraction distributions in the rod bundles were quantified by the high-energy X-ray CT system with a linear accelerator. There are no significant differences in void fraction distributions between condensed sea water and mixture solution. The void-fraction has a peak at the center on horizontal plane for all the fluids. The two salt-waters shift boiling incipience toward downstream and decrease void swell level so that vertical void-fraction profiles of the salt waters are steeper than that of water. This is because created bubbles in the salt waters were smaller than those in water. The spacer has a mixing effect to increase void fraction at upstream of the spacer and homogenize the void fraction on the horizontal plane.

Precipitation profile and dryout concentration of sea-water pool-boiling in 5 × 5 bundle geometry

Abstract As an accident management procedure of light water (nuclear) reactors which are situated along sea shore, sea water will be injected into the reactor pressure vessel to flood the nuclear fuel which is heated by residual heat. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Precipitation behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with condensed (two and half times denser) sea water and a mixture solution of sea water and borated water. Three-dimensional salt-precipitation distributions in the rod bundles were quantified with X-ray CT system. For both solutions, salt precipitated downstream and close to the top of active fuel (TAF) height where the void fraction is the highest. The condensed sea water yields wider precipitation region in height direction than mixture solution does. Mixture solution may give localized precipitates at the same height, which is just below TAF and uniformly spread on the horizontal plane. For both solutions, dryout concentration is larger as collapsed solution level is higher. This is because that lower collapsed solution level gives longer boiling-length and higher void-fraction, which results in larger amount of salt precipitations. The proposed salt concentration is useful to evaluate dryout concentration, which is the almost constant salt concentration for heat flux levels within the experimental ranges.

Development of a method for identifying severe status requiring early entrance of SAMG considering human action time

The margin of human action time from severe accident management guideline (SAMG) entrance to reactor vessel (RV) failure was analyzed to compare the required human action time and available human action time based on a severe accident DB developed using MAAP 5.02 code. In order to analyze the margin of human action time, dominant initiating accidents are selected, such as small break loss of coolant accident (SBLOCA), medium break (MBLOCA), large break (LBLOCA), and station blackout (SBO) which account for about 80 percent of core damage frequency (CDF), and the worst scenario was selected in each initiating accident. In the analysis of the worst scenarios, it was confirmed that a human might be fail to prevent RV failure due to insufficient action time. Therefore, in this study, a method that can determine the SAMG entry condition in advance was developed by applying the concept of component state to function recovery procedure (FRP). To verify the developed method, a case study was performed based on the scenarios that feed & bleed (F&B) and aggressive cool down (ACD) operations are not successful within limited time after entering the FRP. As a result, the developed strategy flow chart can identify the SAMG entry condition in advance, and provide a human with sufficient action time to mitigate severe accidents. It is expected that the results of this paper can provide some insights for revising the SAMG, or for developing an integrated accident management procedures (IAMPs).

Human Error Probability Quantification using SPAR-H Method: Total Loss of Feedwater case study for VVER-1000

The primary side Feed & Bleed (F&B) recovery strategy is the last device to avoid core damage in the event of loss of Ultimate Heat Sink (UHS) accident. In the context of accident management, besides multiple safety systems’ well-functioning, human factor has a significant role in avoiding the extreme scenarios. In this research, it is tried to quantify Human Error Probability (HEP) of operators’ interference to initiate F&B recovery procedure when the Total Loss of Feedwater (TLFW) accident occurs in a typical VVER-1000 Nuclear Power Plant (NPP) via Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) Method. Among those factors that affect human performance is the available time for the operators’ late interference in order to reestablish natural circulation to remove the residual heat after the reactor scrammed which is computed using the best estimate code RELAP5 Mod 3.2. It is revealed that the diagnosis and the action HEPs of the operators are highly dependent and the combined HEP value is too risky with an unacceptable amount of uncertainty. It is concluded that the accident management procedure must be redesigned so as to eliminate diagnosis and action dependency.

Analytical measurements of fission products during a severe nuclear accident

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

The safety analysis of a small pressurized water reactor utilizing fully ceramic microencapsulated fuel

For post-Fukushima nuclear power plants, accident-tolerant fuel (ATF) is strongly desired in order to prevent radioactive release in the event of a severe accident. The Fully Ceramic Microencapsulated (FCM) fuel is one of these materials. The neutronics and thermal hydraulics analysis of small pressurized water reactor (PWR) utilizing FCM fuels needs to be carried out to demonstrate this fuel performance. In this study, first, a three dimensional full core model of a small PWR design is developed to assess the cycle lengths and power distribution. In order to achieve the acceptable cycle lengths, some design optimizations are taken, including higher TRISO particle packing fraction, higher enrichment and using higher density fuels. The results demonstrate that the acceptable cycle lengths can be achieved after this design optimizations. The results reveal that the cycle lengths can be improved to about 550days after using the 19.9w/o enrichment fuels with UO2 fuel kernel, and the cycle lengths can be maintain about 800days with UN fuel kernel due to the higher density than UO2 fuel kernel. Then, suppressing the excess reactivity, the core power distributions and reactivity coefficients are performed to prepare for thermal hydraulics analysis.The steady state and transient state analysis for LOCA is carried out using the best estimate system code RELAP5. For the conventional UO2 fuel, fission product releases simultaneously when the cladding fails. But for the FCM fuel, the radioactive material can be contained in the fuel pellet even after the cladding failure since enveloping the TRISO particles within a dense SiC matrix provides multiple barriers to fission product release. Consequently, although the Zircalloy cladding fails, there is still 3500s available for operator actions before the start of FCM fuel melting occurs even through the SB-LOCA occurring without ECCS, and there is still 1443s available even through the LB-LOCA occurring without ECCS. The results of the accident analysis without ECCS would be useful in understanding cladding and fuel failure progression and could provide guidelines in establishing detailed accident management procedures to design the proper ECCS in the future.

Experimental and theoretical study of large scale debris bed reflood in the PEARL facility

During a severe accident in a nuclear power plant, the degradation of fuel rods and melting of materials lead to the accumulation of core materials, which are commonly, called “debris beds”. To stop core degradation and avoid the reactor vessel rupture, the main accident management procedure consists in injecting water. In the case of debris bed, the reflooding models used for Loss of Coolant Accident are not applicable. The IRSN has launched an experimental program on debris bed reflooding to develop new models and to validate severe accident codes. The PEARL facility has been designed to perform, for the first time, the reflooding of large scale debris bed (Ø540mm, h=500mm and 500kg of steel debris) in a pressurized containment. The bed is heated by means of an induction system. A specific instrumentation has been developed to measure the debris bed temperature, pressure drop inside the bed and the steam flow rate during the reflooding. In this paper, the results of the first integral reflooding tests performed in the PEARL facility at atmospheric pressure up to 700°C are presented. Focus is made on the quench front propagation and on the steam flow rate during reflooding. The effect of water injection flow rate, debris initial temperature and residual power are also discussed. Finally, an analytical model providing the steam flow rate and the quench front velocity is proposed to interpret these results.

Experimental investigation on single-phase pressure losses in nuclear debris beds: Identification of flow regimes and effective diameter

During a severe nuclear power plant accident, the degradation of the reactor core can lead to the formation of debris beds. The main accident management procedure consists in injecting water inside the reactor vessel. Nevertheless, large uncertainties remain regarding the coolability of such debris beds. Motivated by the reduction of these uncertainties, experiments have been conducted on the CALIDE facility in order to investigate single-phase pressure losses in representative debris beds. In this paper, these results are presented and analyzed in order to identify a simple single-phase flow pressure loss correlation for debris-bed-like particle beds in reflooding conditions, which cover Darcean to Weakly Turbulent flow regimes.The first part of this work is dedicated to study macro-scale pressure losses generated by debris-bed-like particle beds, i.e., high sphericity (>80%) particle beds with relatively small size dispersion (from 1mm to 10mm). A Darcy–Forchheimer law, involving the sum of a linear term and a quadratic deviation, with respect to filtration velocity, has been found to be relevant to describe this behavior in Darcy, Strong Inertial and Weak Turbulent regimes. It has also been observed that, in a restricted domain (Re=15 to Re=30) between Darcy and Weak Inertial regimes, deviation is better described by a cubic term, which corresponds to the so-called Weak Inertial regime.The second part of this work aims at identifying expressions for coefficients of linear and quadratic terms in Darcy–Forchheimer law, in order to obtain a predictive correlation. In the case of monodisperse beds, and according to the Ergun equation, they depend on the porosity of the medium, empirical constants and the diameter of the particles. Applicability of the Ergun equation for debris-bed-like particle beds has been investigated by assessing the possibility to evaluate equivalent diameters, i.e., characteristic length allowing correct predictions of linear and quadratic terms by the Ergun equation. It has been observed that the Sauter diameter of particles allows a very precise prediction of the linear term, by less than 10% in most cases, while the quadratic term can be predicted using the product of the Sauter diameter and a sphericity coefficient as an equivalent diameter, by about 15%.

Modelling guidelines for core exit temperature simulations with system codes

Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT).The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects.The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

Assessment of accident management measures on early in-vessel station blackout sequence at VVER-1000 pressurized water reactors

In the process of elaboration and evaluation of severe accident management guidelines, the assessment of the accident management measures and procedures plays an important role. This paper investigates the early in-vessel phase accident progression of a hypothetical station blackout scenario for a generic VVER-1000 pressurized water reactor. The study focuses on the following accident management measures: primary side depressurization with passive safety systems injection, secondary side depressurization with passive feeding from the feedwater system, and a combination of the both procedures. The analyses have been done with the mechanistic computer code ATHLET. The simulations give in-depth analyses of the reactor system behaviour, assessment of the time margins till heating up of the reactor core and insights into physical phenomena which can influence the passive feeding procedures for cooling of the reactor core. The simulation results show that such accident management measures can significantly prolong the time till core degradation. Maximum delay for core heat up can be achieved by sequentially realization of the secondary and primary side bleed and feed strategies. Due to reversed heat transfer in the steam generators or caused by the depressurization itself a part of the injected water is evaporated. Evaporation or flashing in the feedwater system can lead to an intermittent water injection, thus reducing the effectiveness of the feeding procedure.

Travel Time in Case of Accident Prediction Model

The aim of this paper is to estimate the travel time along a motorway section, both in undisturbed condition and taking into account a perturbation due to an accident. The estimated travel time is calculated adding the travel time under normal traffic conditions, the time used settling the accident area and the time needed to dissolve the queue. To be applied the model requires the knowledge of accident characteristics, the traffic flow at the accident time and the value of the reduced capacity during the accident management. The model refers to traffic conditions and accident management procedures typical of motorways.

Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project “TACIS 2.03/97”, Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.