Development of a novel 1D coupled neutronics/thermal-hydraulics code and its verification on PWR rod ejection accident benchmark

This paper presents a new 1D Neutronics/Thermal-hydraulics code ATAC-1D based on the advanced Jacobian-Free Newton-Krylov (JFNK) method and the low dimensional equivalent strategy. Conventional operator-splitting (OS) strategies are used to maintain its accuracy with small time steps and linearization of the nonlinear problem, which leads to slow computation speed and linearization error. The JFNK method solves the troubles in the coupled neutronics/thermal-hydraulics problems mentioned above. Furthermore, a core-wide three dimension to one dimension equivalent method has been developed to provide variable few-group parameters. Finally, the performance of the coupled neutronics/thermal-hydraulics code ATAC-1D is studied by simulating four OECD/NEA CRP PWR rod ejection benchmark problems. The simulation results are compared to the reference ones, which proves that the developed 1D code has a good accuracy and practicability in nuclear reactor transient calculation.     

Experimental verification of accident loads for PWR containments

The principal loads which a nuclear power reactor containment is designed to withstand are produced by internal fluid caused static and/or dynamic pressures. They can be generated during failure events which release mass and energy into the containment atmosphere. An overview of the events which can generate substantial internal loads is provided. Representative experimental programs initiated for the investigation of the relevant physical phenomena are described. Illustrative examples of measured data are presented and discussed.     

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.     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

The study on a statistical methodology for PWR fuel rod internal pressure evaluation

The most limiting design criteria for high Burnup PWR fuel are known to be rod internal pressure and cladding oxidation. Some fuel vendors have been increasing the design margin of rod internal pressure by increasing fuel rod plenum volume or optimizing fuel pellet grain size. In this study, a sophisticated statistical methodology that employs the response surface method and Monte Carlo simulation has been proposed to increase the design margin of rod internal pressure and subsequently a simplified statistical methodology has been developed to simplify the sophisticated statistical methodology. The simplified statistical methodology utilizes the system moment method combined with a deterministic approach for calculating a maximum variance of rod internal pressure. This simplified statistical methodology may be more efficient in the reload core fuel rod performance analyses than the sophisticated statistical methodology since the former eliminates numerous calculations needed for the evaluation of power history-dependent variances. It is found that this simplified methodology also generates more conservative rod internal pressure than the typical statistical methodology.     

Procedure for testing VVÉR-1000 fuel in the MIR reactor under the conditions of an accident with ejection of a control rod

A procedure for performing dynamical experiments to study the behavior of VVéR fuel under the conditions of an anticipated accident with the introduction of positive reactivity has been developed and adopted on the MIR reactor. A power pulse is produced by removing the screen from the absorber in the initial state of the surrounding fuel element. A reactivity compensator, which is moved synchronously with the screen, is provided to decrease the effect on the core. A three-element fragment of a VVéR-1000 fuel assembly is used as an experimental fuel assembly. The volume of the installed measuring sensors and the rate at which the parameters are recorded are adequate for calculating the thermomechanical state of a fuel element. The precedure developed makes it possible to obtain in an experiment the lineal power density of fuel elements and the pulse parameters that correspond to the computed values for an anticipated accident. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 427–431, December, 2006.     

Experimental study of jet discharging test results under BWR and PWR loss of coolant accident conditions

This report describes the results of the jet discharging experiments conducted at the Japan Atomic Energy Research Institute. The tests were done under BWR and PWR Loss of Coolant Accident conditions using 4 inch, 6 inch and 8 inch test pipes, and varying distance between the pipe exit and the target plate.Simple and practical experimental formulae to estimate the maximum pressure on the target plate and maximum pressure distribution are given. Further, relations between pipe reaction thrust forces and jet impingement forces are described.     

Investigation of PWR accident situations at BETHSY facility

For several years an extensive programme of separate effect tests related to PWR safety has been conducted in France and in particular at the Nuclear Center of Grenoble. Recently the BETHSY integral test facility - three identical loops, full height and pressure - has been constructed with the main objectives of contributing to the verification of the CATHARE calculation code of accidents and to the validation of the physical bases of Emergency Operating Procedures.So far several tests have been (November 1988) carried out, among which natural circulation under various conditions (single phase, two-phase, symmetric or asymmetric conditions, different core power, variable steam generator liquid level), 2 inches cold leg break, steam bubble formation and collapse in the upper head. Summarized results of some of these tests are introduced and compared to a first CATHARE calculation as far as the 2″ cold leg break is concerned.     

严重事故下大功率先进压水堆IVR-ERVC有效性分析

通过压力容器外部冷却(ERVC)以实现堆内熔融物滞留(IVR)作为反应堆严重事故缓解管理的一项重要举措一直以来广泛受到关注和研究。本文使用严重事故分析程序MELCOR,从瞬态角度对大型先进压水堆进行了IVR-ERVC相关研究。过程中重点关注了堆芯熔毁和重新定位,熔池形成、生长及其传热过程,并且对压力容器外部流动传热进行了分析。MELCOR计算所得下封头热流密度分布的瞬态结果与临界热流密度(CHF)比较和分析表明,1700 MWe大功率压水堆发生严重事故后在IVRERVC条件下能够保证压力容器的完整性,即,IVR-ERVC能够有效带出下封头熔融物的衰变热量,缓解严重事故后果。     

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines （SAMG）, it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident （LOCA）, Steam Generator Tube Rupture （SGTR）, Station Blackout （SBO）, and Anticipated Transients without Scram （ATWS） in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management （SAM） such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.     

Severe accident analysis of a 1300 MWe PWR to address the hydrogen issue

The behaviour of the potentially large quantity of hydrogen generated during a severe accident has been recognised as an issue of importance since the accident at Three Mile Island. In this article, we describe a severe accident analysis for the Neckarwestheim 2 1300 MWe PWR “Konvoi” plant, performed primarily to investigate the behaviour of hydrogen in the containment, and draw conclusions regarding the need for hydrogen control systems (igniters). The Modular Accident Analysis Program (MAAP) developed by IDCOR in the United States, and the Westinghouse COMPACT multi-compartment containment code were used. The study investigated the generation, release to containment, distribution within containment and potential combustion of hydrogen produced during two severe accident sequences. Results are summarized which show that hydrogen mixing in containment is generally good and that even without hydrogen control systems, hydrogen combustion, although possible, does not threaten containment integrity.     

Cobalt irradiation box ejection accident of ETRR-2

The new Egyptian Test and Research Reactor Number 2 ETRR-2, MTR type, is now under operational tests. It has a main central irradiation channel for the purpose of Co⁶⁰ isotope production with an intended rated capacity of 50 000 Ci per year. The reactivity introduced in the reactor due to accidental ejection of the Co⁶⁰ irradiation box (CIB) should be discussed. This reactivity insertion accident (RIA) may be fast or slow with maximum reactivity worth 2.9428 $. The CIB may move with constant speed or variable acceleration according to its initial speed and the applied forces. This results in a linear, parabolic or sinusoidal motion, which in turn affects the reactivity insertion rate (RIR). The present work analyzes this type of perturbation during normal operating conditions: 22 MW full power and 1900 kg s⁻¹ forced core cooling flow. The work serves as a part of the safety evaluation process applicable to similar MTR cores. The RIA code TRANSP20 is developed for this study. It simulates various types of RIR, fast or slow resulting from different CIB ejections. Scram signal due to power, period, inlet and outlet temperatures, or temperature difference is expected to activate the shutdown system. The work presents five case studies, two for fast ejection and three for slow. The transient behavior of the reactor during this is illustrated. The results show that the reactor can withstand slow ejection if the scram is available. However, for fast ejection the scram system does not prevent the clad temperature from exceeding safety limits. Recommendations to prevent or mitigate this accident are highlighted.     

Theoretical predictions of failure probabilities for PWR pressure vessels subjected to accident conditions

The Second Marshall Report (1982) presented a detailed analysis of the integrity of PWR pressure vessels. As part of that study theoretical calculations of failure probabilities were made. Since the publication of that Report modifications have been made to the theoretical model to extend the failure criterion into stable crack extension, to update the knowledge of the distribution of various parameters, to more accurately represent the stress intensities and crack shapes, and to consider a different representation of pre-service detection of defects.In this paper these modifications are summarised and the application of the model to the calculation of failure frequencies for the most severe accident conditions, the large loss-of-coolant accident and the steam break, is presented. The results indicate that the probability of vessel failure is one to two orders of magnitude lower than previously predicted.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

Methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly in a PWR

In a pressurised water reactor, the rod cluster control assembly is a system which controls the neutronic activity of the core. It consists of long rods, connected by a spider fixture and a cylindrical system for the control drive mechanism. At its withdrawn position, the activity of the core is maximum, and at its completely inserted position, the activity of the core vanishes. In case of emergency, an effective way to shutdown the reactor is to let it drop under its own weight. An other way to verify the efficiency of the rod cluster control assembly is the insertion test. It consists in inserting the rod into its guides and in checking if the reaction friction force is not high enough to block the movement of the rod cluster control assembly.We present in this paper a methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly into its guides (discontinuous and continuous guides, guide thimble). A numerical model is elaborated in which many loads are taken into account: fluid load, gravity and friction force between the rod and the guide. The numerical results are compared to experimental measurements obtained from a full-scale structure. A good agreement between the calculated and the measured data is observed.The numerical model takes into account the possible deflection of the guide. It shows clearly that the friction force cannot be neglected when the guide is bowed. So one can locate a faulty guiding system by examining the reaction force during the insertion test. Then, the numerical model can help the decider to make his choice among different rod cluster/fuel assembly components.     

A theoretically based critical heat flux prediction for rod bundles at PWR conditions

A theoretically based procedure developed for round tubes has been applied to the prediction of DNB heat fluxes in rod bundles at PWR conditions. State-of-the-art subchannel analysis procedures were used to determine local flows and enthalpies. Very good comparison between DNB predictions and experimental observations are found for rod bundles which both uniform and non-uniform axial heat fluxes.     

Characterization of PWR vessel steel tearing under severe accident condition temperatures

In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment.On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance.The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens.The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead to different types of rupture behaviour at high temperatures.The experimental programme has been conducted on various French RPV 16MND5 steels for temperatures ranging from 900 °C to 1100 °C. Comparisons between the tests performed on these various 16MND5 steels show that this approach is appropriate to characterize the difference in ductility observed at high temperatures.The aim of this experimental study is also to contribute to the definition of a tearing criterion by identifying, on the basis of CT results, the related material parameters at temperatures representative of the real severe accident conditions.This experimental campaign has been carried out in partnership with IRSN in the framework of a research programme whose purpose is to complete the mechanical properties database of 16MND5 steel and to model tearing failure in French RPV lower head vessels under severe conditions (Koundy et al., 2008).     

Analysis of flow distribution a PWR fuel rod bundle model containng A 90% blockage

An experimental investigation, covering a Reynolds number range from 2 × 10³ to 3.5 × 10⁴, was conducted to study the velocity and turbulence intensity distributions due to the presence of a blockage in an unheated 7 × 7 rod bundle. The blockage configuration, consisting of a 4 × 4 rod array, created a maximum flow area reduction of 90% in the central nine subchannels. The blockage sleeve length was 38.3 × rod diameter and the 90% blockage zone length extended for 16.4 × rod diameter. The results showed that upstream of the blockage, the flow was not influenced by the blockage until it reached the location where the inlet taper section of the swelling started. At the downstream end, the flow disturbance was extensive and persisted over a distance of about 83 rod diameters. Compared to the downstream velocity profiles, the turbulence intensity measurements however showed a faster recovery from the blockage influence. At the higher Reynolds number, velocity profiles calculated using the COBRA subchannel computer code compared consistently with the experimental data. The general flow behaviour of the various subchannels was reasonably well predicted. However, at low Reynolds number, due mainly to the frictional form loss calculation scheme in COBRA and uncertainty in the flow transition, the flow diversion due to the blockage to the surrounding unblocked subchannels was overestimated. The influence of the degree of recovery from the rod swelling on the flow was also studied using COBRA.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity.