A procedure to optimize the timing of operator actions of accident management procedures

The analysis of beyond design basis accidents (BDBA) is an essential component of the safety concept of nuclear power plants (NPP). Goal of the analysis is to achieve a set of actions aimed to prevent the escalation into a severe accident, to mitigate consequences of a severe accident, and to achieve a long term controllable state of the NPP. This paper presents an analytical procedure to optimize the timing of operator interventions. The procedure is demonstrated based on four sets of parameters, first, parameters which define the operator actions are chosen. Second, parameters which define the system availability are chosen. Third, parameters which define in a continuous way the status of the plant are chosen. Finally, one looks for a functional dependency of the accident management (AM)-parameters and the parameters describing the plant status. Once a function could be found, this function is “optimized” in the sense that the AM-parameters are varied to find a optimal overall condition for the plant. In the first part, the paper presents the analytical procedure in a general way, in the second part, an initiating event is chosen. The procedure is applied to a station black out (SBO) transient, and as operator action secondary side bleed and feed, followed by primary side bleed and feed, is foreseen. As result, the optimal timing to initiate both actions is achieved.     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

Analysis of the degradation of a PWR control rod during a loss of coolant accident, using the ICARE2 code

Conditions leading to AIC control rod damage during a loss of coolant accident in a PWR geometry, even in absence of violation of the LOCA licensing criteria, are investigated using several versions of the ICARE2 code (IPSN). Before being applied to the reactor case, the code and the modelling procedure are validated against the out-of-pile severe fuel damage experiment CORA-5. Three particular initial configurations are considered for the subsequent control rod damage analysis: nominal control rod and guide tube geometry, zircaloy guide tube bowing with concurrent cladding thickness reduction and finally control rod cladding perforation. For each of these cases the thermal, mechanical and chemical behaviour is presented. Phenomena such as ballooning and cladding failure of fuel rods, guide tube failure, melt relocation and final fluid channel cross-section modification are described. Finally, the conclusions of numerous sensitivity studies are discussed and some suggestions are given for possible improvements of the ICARE2 code.     

Analysis of loss of flow accident at Pakistan research reactor-1

The main objective of the reactor safety is to keep the reactor core in a condition, which does not permit any release of radioactivity into the environment. In order to ensure this, the reactor must have sufficient safety margins during all possible operational conditions (normal as well as accidental). To accomplish this, a study has been carried out, for the analysis of loss of flow accident (LOFA), which is one of the probable scenarios among other possible events such as reactivity-induced-accidents, loss of coolant accident, etc. The study has been carried out for Pakistan research reactor, PARR-1, which was initially converted from HEU to LEU fuel. It is a swimming pool type reactor using MTR type fuel. Presently, a new core is proposed to be assembled containing LEU and some of the used (less burnt) HEU fuel elements. The accident is assumed when the reactor is running at a steady-state power level of 9.8 MW. Computer code PARET and standard correlations were employed to compute various parameters. Results predict nucleate boiling in the core but the temperatures would remain far below the fuel clad melting point.     

Evaluation of heat transfer by sublimation for the application to the divertor heat sink for high fusion energy conversion

Thermal and structural responses of divertor target were evaluated by using finite element method. High heat flux simulating ELMs at the level of 100 MW/m² was assumed onto the tungsten armor, and surface temperature profile was obtained. When dynamic heat load over 100 MW/m² was applied, the maximum surface temperature exceeded 1300 °C, and it caused recrystallization of tungsten regardless of the heat transfer below it. The result was used to conduct dynamic heat load experiment on tungsten, and material behavior of tungsten was evaluated under dynamic heat load. This study also proposed new concept of divertor heat sink which can distribute high heat flux and transfers the heat to high temperature medium. It consists of tungsten armor, composite enhanced with high thermal conductivity fiber, and heat transport system applying phase transition. High heat flux simulating ELMs was also applied to target surface of the divertor, temperature gradient, thermal stress of tungsten and composite were evaluated. Based on the results of analysis, thermal structural requirement was considered.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

Characteristic features of the legal relations concerning compensation for injuries caused by the chernobyl accident

This article is devoted to the legal relations between the State and citizens who were victims of the accident at the Chernobyl nuclear power plant. Their characteristic features consist in the fact that the government as the constitutional guarantor and owner of nuclear power plants is responsible for the harm done due to radiation. The method of compensating for radiation damage in the form of compensations and benefits for harm to property and health of the victims was not known to the acting legislature before April 26, 1986. Compensations and benefits are also given to categories of citizens who are healthy and capable of working but were subjected to irradiation for risk of possible radiation-induced injury appearing in the future. 5 references. Concern Rosénergoatom. Translated from Atomnaya énergiya, Vol. 87, No. 1, pp. 68–71, July, 1999.     

Application of risk-informed design methods to select the PSACS ultimate heat sink

In the early phases of advanced system design, information is scarce. The technologies, components and processes to be used have not been specified adequately or are not well understood and uncertainties are very large. Yet, it is during these early phases that design teams and other stakeholders are required to make critical decisions to guide the development of the system. To aid in this decision making, a formal process is proposed based on the Analytic-Deliberative Decision-Making Process (ADP) that allows stakeholders to synthesize rationally their knowledge and experience and facilitate learning and sharing of best practices. The ADP identifies and prioritizes attributes relevant to a decision problem and supports the formulation of metrics to measure the performance of different design options. This paper reports on an application of the ADP to the selection of an ultimate heat sink for the Flexible Conversion Ratio (FCR) reactor's Passive Secondary Auxiliary Cooling System (PSACS). Two ultimate heat sink options are identified and evaluated, air and water.     

A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.     

Modeling the heat transfer during an accident with sharp reactivity increase in a water-cooled,water-moderated reactor

Moscow Power Institute. Translated from Atomnaya Énergiya, Vol. 74, No. 3, pp. 199–210, March, 1993.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

Analysis of heat and mass transfer to determine heat loss and the rate of condensation of the MVSTs off-gas ducts

Reduction of the existing nuclear waste in the Melton Valley Storage Tanks (MVSTs) at the Oak Ridge National Laboratory (ORNL) is of utmost concern to the scientists at this facility. This paper provides proof that a combination of vault heating, sparged air heating, and prevention of condensation is the best alternative to achieve this goal. Therefore, in this study a general system of mathematical equations has been developed taking into account all of the parameters affecting evaporation and condensation. This evaporation process has been analyzed by the careful modeling of a bubble chain through the extremely viscous, radioactive liquid contained in the storage tanks. This paper discusses in detail the evaporation procedure using bubble formation, air velocity, and determining the rate at which this liquid waste can be removed from the MVSTs by evaporation under different conditions of the sparging air. An additional objective is to study the heating/cooling of the condensation process of the off-gas piping inside the vault. A laboratory scale model has also been assembled for this purpose at ORNL to verify the accuracy of the mathematical modeling. A comparison of the experimental findings with the mathematical modeling shows excellent agreement.     

Analysis of the cooling capability of steam generators during the loss of residual heat removal system at midloop operation with closed primary system

Loss of residual heat removal system (RHRS) at midloop operation is one of the most significant core damage risk contributor at low power and shutdown conditions. During this kind of transients the reflux-condensation is one of the cooling mechanisms anticipated in the abnormal procedure of loss of RHRS at midloop level. In this sense, several simulations of loss of the RHRS with closed primary system with the TRACE V4.160 code have been performed considering different availability of steam generators. The present study aims to analyze the thermal-hydraulic behavior after the loss of RHRS at midloop conditions with the reflux-condensation as the only cooling mechanism available and to investigate the capability of this cooling mechanism. The simulation results show that one steam generator is sufficient to remove core decay heat of 11 MW obtaining an equilibrium pressure, but the core uncovery depends on the number of steam generators operating. Finally, an analysis of the abnormal procedure and the event trees of the loss of RHRS sequences at midloop operation has been performed taking into account the results obtained in the simulation with TRACE.     

Effects of oxide and hydrogen on the behavior of Zircaloy-4 cladding during the loss of the coolant accident (LOCA)

Effects of surface oxide and absorbed hydrogen on the behavior of the loss of the coolant accident (LOCA) were investigated in this study. High temperature ballooning and thermal quench tests were performed for Zircaloy-4 cladding which had been prepared with up to 50 μm of oxide and 1000 ppm of hydrogen, respectively. In the high temperature ballooning test, the initially pressurized cladding was heated until a rupture. Threshold oxidation (ECR) of each condition was evaluated in the thermal quench test in which oxidized cladding at the LOCA temperature was quenched by water. Ring compression test was performed to assess the ductility of the quenched cladding The results showed that both the oxide and hydrogen affected the high temperature ballooning property due to the constraint of the α phase by the surface oxide as well as the expansion of the β phase by the absorbed hydrogen. In the quench test, the pre-oxide and absorbed hydrogen did not affect the high temperature oxidation whereas the threshold ECR decreased in the hydrogen charged cladding because the absorbed hydrogen increased the maximum oxygen solubility inside the residual β layer to reduce the cladding ductility.     

Simulation of a small loss of coolant accident by using RETINA V1.0D code

RETINA has been developed for modeling of two-phase flow situations in full-scope simulators of nuclear power plants. A special feature of RETINA is that both RETINA V1.0D (drift-flux — 5 equations) and RETINA V1.0-2V (two-fluid — 6 equations) approach are available in the code and the same constitutive relations are used in both cases. The governing equations are discretized implicitly, and an automatic derivation algorithm determines the Jacobian matrix, which is partitioned taking into account the special structure of nuclear power plants. Partitioned inverse formula is used to solve the global equation system providing the possibility of multi-level parallelization. Heat structures are modeled in two dimensions and are coupled to the flow equations explicitly. Since the code will be used in real-time simulators, we paid special attention to time-effective solution. In this paper, we demonstrate the ability of our code by simulating a small loss of coolant accident Paks Model Circuit (PMK). The simulation results are compared to real measurements obtained by Paks Model Circuit.     

Analysis of a loss of residual heat removal system during mid-loop conditions at PKL facility using RELAP5/Mod3.3

When a nuclear power plant is in shutdown conditions for refuelling, the reactor coolant system water level is reduced. This situation is known as mid-loop operation, and the residual heat removal (RHR) system is used in this situation to remove the decay power heat generated in the reactor core.In mid-loop conditions, some accidental situations may occur with not a negligible contribution to the plant risk, and all involve the loss of the RHR system. Thus, to better understand the thermal–hydraulic processes following the loss of the RHR during shutdown, transients of this kind have been simulated using best-estimate codes in different integral test facilities. This paper focuses on the simulation, using the best estimate code RELAP5/Mod3.3, of the experiment E3.1 conducted at the PKL facility. This experiment consists of the loss of the RHR system when the plant is in mid-loop conditions for refuelling and with the primary circuit closed. In particular, in this experiment the physical phenomena to investigate are the mechanisms of heat removal in presence of nitrogen and the deboration in critical parts of the primary system.