Design Basis Accident Conditions Research Articles

CrN–NbN nanolayered coatings for enhanced accident tolerant fuels in BWR

The accident tolerant fuel (ATF) concept has emerged in the years after the 2011 Fukushima accident as part of a renewed effort in research for light water reactors. The primary focus is to further improve safety measures under and beyond design basis accident conditions, and to improve fuel cladding performance in normal operation. The application of a coating on zirconium claddings can achieve both these aims without extensive changes to the reactor design. Metallic chromium coatings have been profusely studied as solution for pressurized water reactors, but the search for an effective ATF coating able to withstand the environment inside boiling water reactors (BWRs) is still ongoing. In this work, two different versions of a novel nitride coating composition were studied. Zirconium claddings coated with 8 µm thick layers of superlattice CrN–NbN and a nanolayered CrN–NbN were tested in autoclave under BWR operating conditions for 60 days. Scanning electron microscopy, transmission electron microscopy, energy dispersive x-ray spectroscopy, electron back-scattered diffraction, x-ray diffraction, and atom probe tomography were employed to characterize as-deposited and autoclaved samples of these two materials. During exposure, both coating versions formed a stable, dense and passivating oxide scale (200–300 nm thick) on the surface, demonstrating improved oxidation protection under operating conditions. Some differences in the oxide growth mechanism were observed between the superlattice and the nanolayered CrN–NbN coatings, which allowed to glimpse at the effect of the layer thickness on the oxidation protection provided by these coatings. The nano-structured morphology of both coatings remained unaffected by the autoclave test, but a 35 nm thick Zr-Cr-N phase was found at the coating-substrate interface of the superlattice CrN–NbN coated cladding.

Analysis of BDBA sequences in a generic IRIS reactor using ASTEC code

Integral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal–hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs.

THE VALIDATION OF COMPREHENSIVE SOFTWARE FOR THE ASSESSMENT OF RADIOLOGICAL IMPACTS ON HUMANS THROUGH BENCHMARK ANALYSIS WITH RADIOLOGICAL CODES.

In this paper, a novel domestic code called KIANA was developed for the assessment of radiological impacts on the population in normal and accident conditions including design basis accident (DBA) and beyond DBA (BDBA) for the nuclear power plants. The validation process of the KIANA code was performed using the results of the DOZA_M radiological code, whose results are presented in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant Unit One (BNPP-1). The calculations of KIANA are performed based on the Gaussian diffusion model. The developed KIANA code has the potential of calculating the concentration and radionuclide doses due to the pathways such as airborne, foodstuff, marine (both one and two boxes models), soils, animals, vegetation (with and without tritium) and other pathways without any restriction. In the current research, the individual dose from a cloud to the member of the public in the region of BNPP-1 in normal condition was calculated. Moreover, the total effective dose to the member of the public from the primary to the secondary leakage inside steam generators, large break loss-of-coolant accident (LBLOCA) and small break loss-of-coolant accident was calculated. Thyroid gland equivalent dose for the infant (1-8 years) in the case of LBLOCA at the BNPP in DBA conditions was also evaluated. Finally, the prevented dose at the initial stage for the whole body of adults after BDBA, prevented dose at the initial stage for the thyroid gland of children after BDBA and the effective dose during the first year after the accident (external body irradiation from presence in the area) in the case of BDBA are assessed. The KIANA simulation results showed a good agreement with the FSAR data of BNPP.

Characterization of Kanthal APMT and T91 oxidation at beyond design-basis accident temperatures

Limited information is available on the oxidation mechanism of accident tolerant claddings (ATC) Kanthal APMT and T91 at the onset of beyond design-basis accident (BDBA) conditions. We characterized the surface of these ATC alloys after steam and air exposure at 1200 °C for 2 h, defining the oxidation mechanism. Thickness and composition were analyzed with microscopy, Raman spectroscopy, and synchrotron diffraction. Our results demonstrate that APMT forms a compact and homogeneous α-Al2O3 layer when exposed to air or steam. T91 forms a heterogeneous porous layer, containing a mixture of Cr- and Fe-based oxides, whose composition changes with the exposure environment.

Unveiling damage mechanisms of chromium-coated zirconium-based fuel claddings by coupling digital image correlation and acoustic emission

Coated nuclear fuel claddings offer a promising, near-term solution to address the demand for next-generation, accident-tolerant fuel systems. It is expected that they will possess superior mechanical properties and greater oxidation resistance compared to current cladding technology, allowing for improved performance during beyond design-basis accident conditions. Here, we present a methodology to determine the failure mechanisms of chromium-coated zirconium-alloy claddings under expected, critical loading conditions. Three-dimensional digital image correlation and acoustic emission techniques were used in situ to monitor spatial strain and crack development of the claddings under two key, de-coupled loading conditions: expanding plug and four-point bending. Critical strain levels, at which cracking initiated, were determined to be 0.4% εhoop and 0.4% εxx for expanding plug and four-point bending, respectively. A two-dimensional fracture model was also developed for the expanding plug loading condition based on inputs determined from mechanical testing. It was concluded that coating fracture of expanding plug specimens was axisymmetric across the specimen circumference and occurred rapidly through the thickness of the chromium coating. Subsequent high temperature steam oxidation experiments on tested (cracked) expanding plug specimens showed no signs of oxidation degradation to the underlying zirconium alloy, thereby showcasing the effectiveness of the chromium coating. This comprehensive, multi-scale study is intended to inform future testing of next-generation, coated claddings and identify the resulting failure mechanisms that arise in beyond design basis accident conditions.

Study on dryout and rewetting during accidents including ATWS for the BWR at JAEA

Post-boiling transition (Post-BT) heat transfer is important to analyze the duration of surface dryout and the peak cladding temperature during abnormal transient and design basis accident (DBA) conditions in light water reactors. Furthermore, the Japanese new safety regulation revised after the Fukushima Daiichi accident requires the analysis of the beyond-DBA conditions including the anticipated transient without scram (ATWS), where experimental databases for the code validation are almost-non-existent especially for those with highly oscillatory power change caused by thermo-nuclear coupling. Motivated by the importance, Japan Atomic Energy Agency has conducted a series of experiments for BT and rewetting phenomena to develop mechanistic models based on physical understanding and to validate thermal hydraulic codes. So far, the experimental databases for the Post-BT heat transfer rate, deposition rates of liquid droplets, and the dryout and rewetting behavior were developed mainly for the conditions covering the anticipated operational occurrences (AOO) for BWRs. The significance of the precursory cooling just before the rewetting was identified from comparison with a theoretical model for the two-dimensional transient heat conduction in the cladding. Recently, experiments including the simulation of the ATWS were initiated using the single-tube test apparatus with simulated spacers, liquid film heat transfer visualization apparatus for visual observation, and the HIDRA facility with test sections of a full-length 4 × 4 bundle and a short-length 3 × 3 length bundle. Some promising results have already been obtained, which show effects of simulated ferrule and swirl spacers on the post-BT heat transfer and rewetting velocity, and mechanism of the rewetting by the propagation of the liquid film on the initially heated surface of the heater rod.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors

Coatings with thicknesses between a few microns and ∼10 μm deposited on a Zircaloy-4 substrate have been studied with the objective to provide a significant reduction in the oxidation-induced embrittlement of the nuclear fuel cladding, especially in accidental conditions, such as LOss-of-Coolant-Accident (LOCA) conditions. This paper deals with the early studies carried out at CEA, several years before the Fukushima-Daiishi events, on different types of coatings obtained by a physical vapor deposition process. The studied coatings included ceramic, nitride and metallic multi-layered ones. The results of this screening analysis showed that the first generation of chromium-based coatings exhibited the most promising behavior: good compromise between oxidation resistance and adhesion to the metallic substrate, good fretting resistance and improved resistance to oxidation in steam at high temperature (Design Basis Accident LOCA conditions and slightly beyond).

Turbopump scaling analysis and similarity level estimation for Texas A&M university RCIC system experimental test facility

The Reactor Core Isolation Cooling (RCIC) System is a safety system that provides water to the reactor pressure vessel during off-normal BWR reactor conditions, such as reactor isolation from the turbines or loss of AC power. Under loss of AC power conditions, the RCIC System is expected to fail due to battery depletion within 4–8 h of operation for many units. However, the system did not fail until about 70 h into the accident at Fukushima Dai-ichi Unit 2, which was well past the time of battery depletion. To investigate the full potential of the RCIC System, the Laboratory for Nuclear Heat Transfer Systems (NHTS) at Texas A&M University is designing and constructing an experimental RCIC System facility. A careful scaling analysis is essential to ensure proper representation of the RCIC System's key components and phenomena in the experimental testing.This paper describes a method to estimate the similarity level (SL) of the RCIC turbine-pump assembly (turbopump), along with application of the methodology using the Peach Bottom Unit 2 RCIC System as the reference full-scale system. The methodology is demonstrated with the Texas A&M University facility but can be applied to other RCIC system facilities.Zuber's H2TS (Hierarchal Two-Tiered Scaling) method has been adopted to specify the required level of detail for the scaling analysis. The analysis includes development of unitless characteristic time ratios to quantify the degree of similarity of the RCIC turbomachinery between the test facility and the prototype. The RCIC turbine and pump are focused on herein because, first, the Fukushima Dai-ichi Unit 2 RCIC System may have ceased operation at 70 h due to turbine failure, and, two, the greatest RCIC System modeling challenges are with the long-term operation of these turbomachinery components. Since the Texas A&M experiments will provide validation data for analytical models of the turbomachinery, applicability of the data must be evaluated.Based on the similarity level estimates, there is a high degree of similarity between the NHTS and Peach Bottom RCIC System turbines, with a minimum similarity level of 75%. This allows the use of the NHTS RCIC turbine for testing over the full range of Design Basis Accident conditions. Another important finding from this study is that the pump configuration initially designed into the NHTS facility had a low similarity level between the experimental and full-scale facilities. This paper identifies pump parameters in the NHTS facility that can be changed to achieve greater similarity. The similarity level analysis indicate that the NHTS RCIC turbine control volume is appropriate for representing the full-size turbomachinery and can be used to study the prototype turbine and pump behaviors under design basis operating conditions.Scaling analyses of the other RCIC System components are to follow the turbomachinery similarity evaluations. The NHTS facility is currently being prepared to test over the entire Design Basis Accident containment pressure and temperature ranges. Future efforts will also focus on modeling, experimentally and with Computational Fluid Dynamics tools, the RCIC System to provide data for beyond design basis operational conditions.

Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination

Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.

Fragility comparison analysis of CPR1000 PWR containment subjected to internal pressure

Evaluating the reliability of the containment structure of a nuclear power plant under an accident internal pressure is of importance in Design Basis Accident and Design Extension Condition analyses. To study the impact of the diverse requirements concerning material properties in various Nuclear Power Plant (NPP) design codes on the reliability of the CPR1000 PWR containment structure, its detailed full-sized three-dimensional Finite Element (FE) model with nonlinear material constitutions, accurate geometric features, and a complex reinforcement system is constructed, and a set of material property data is obtained through practical measurements at a certain CPR1000 power plant for comparison. Based on the Monte Carlo method, random properties of the main material of the containment are applied to produce 100 samples using two sets of NPP design codes (Chinese and European) and one set of practically measured data. Various damage patterns of the samples generated owing to the uncertainty in the material properties are distinguishable. Generally, a concrete crack near an equipment hatch causes containment functional failure and when pre-stressed tendons begin to yield, the containment structure would not continue to maintain its integrity. Based on the functional and structural failure criteria, the corresponding fragility curves are obtained, and the results suggest that irrespective of the differences in the definition of material properties in various NPP design codes, the calculated reliability is practically the same. The difference of internal pressures corresponding to 5% failure probability is less than 2.6%. However, the containment based on practically measured data shows a slightly higher and slightly lower reliability than that based on Chinese and European design codes, respectively.

Neutron cross section sensitivity and uncertainty analysis of candidate accident tolerant fuel concepts

The aftermath of the Tōhoku earthquake and the Fukushima accident has led to a global push to improve the safety of existing light water reactors. A key component of this initiative is the development of nuclear fuel and cladding materials with potentially enhanced accident tolerance, also known as accident-tolerant fuels (ATF). These materials are intended to improve core fuel and cladding integrity under beyond design basis accident conditions while maintaining or enhancing reactor performance and safety characteristics during normal operation. To complement research that has already been carried out to characterize ATF neutronics, the present study provides an initial investigation of the sensitivity and uncertainty of ATF systems responses to nuclear cross section data. ATF concepts incorporate novel materials, including SiC and FeCrAl cladding and high density uranium silicide composite fuels, in turn introducing new cross section sensitivities and uncertainties which may behave differently from traditional fuel and cladding materials.In this paper, we conducted sensitivity and uncertainty analysis using the TSUNAMI-2D sequence of SCALE with infinite lattice models of ATF assemblies. Of all the ATF materials considered, it is found that radiative capture in 56Fe in FeCrAl cladding is the most significant contributor to eigenvalue uncertainty. 56Fe yields significant potential eigenvalue uncertainty associated with its radiative capture cross section; this is by far the largest ATF-specific uncertainty found in these cases, exceeding even those of uranium. We found that while significant new sensitivities indeed arise, the general sensitivity behavior of ATF assemblies does not markedly differ from traditional UO2/zirconium-based fuel/cladding systems, especially with regard to uncertainties associated with uranium.We assessed the similarity of the IPEN/MB-01 reactor benchmark model to application models with FeCrAl cladding. We used TSUNAMI-IP to calculate similarity indices of the application model and IPEN/MB-01 reactor benchmark model. This benchmark was selected for its use of SS304 as a cladding and structural material, with significant 56Fe content. The similarity indices suggest that while many differences in reactor physics arise from differences in design, sensitivity to and behavior of 56Fe absorption is comparable between systems, thus indicating the potential for this benchmark to reduce uncertainties in 56Fe radiative capture cross sections.

System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions

This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.

Application of best estimate thermal hydraulic code for safety analysis of high flux research reactor

Thermal hydraulic safety analysis of nuclear reactors is performed with an objective to ensure that adequate cooling is provided for the reactor core during normal operation, during anticipated transients and also during design basis accident conditions. Though an established international expertise in relation to computational tools, procedures for their application including Best Estimate (BE) methods and comprehensive experimental database exists for Nuclear Power Plants (NPP), this is lacking for research reactors. The importance of transferring these tools and methods of NPP to research reactors has been initiated by IAEA. However, the ranges of parameters of interest in research reactors are quite different from that of NPP due to large variation of their fuel composition, system pressure, materials and overall system configuration. So far, conservative computational tools are used to perform safety analyses for design of these reactors. The application of BE methods has become a real necessity in the safety analysis and design of research reactors. This allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best Estimate methods result in a precise prediction of the system behaviour leading to mitigation of constraining limits in design and operation. The purpose of the present paper is to provide an overview of the thermal hydraulic safety analyses carried out for the design of high flux research reactor using best estimate thermal hydraulic code RELAP5. Various parameters such as fuel and clad temperature distribution, ONB temperature and DNBR under all anticipated/operational transients and design basis accidents are presented.

Experimental results on the breakaway oxidation of the E110 cladding alloy under high-temperature isothermal conditions

In order to evaluate the effect of zirconium breakaway oxidation on the behaviour of nuclear fuel in LOCA (Loss-of-coolant accident) conditions, high temperature (800–1200 °C) oxidation tests with E110 type cladding were performed in steam atmosphere. The onset time of breakaway oxidation was detected using an online method based on hydrogen release. The experimental results showed that the breakaway oxidation starts not earlier than 5 min after the start of the oxidation that is longer than the duration of dry phase in a design basis accident LOCA. The experiments give preliminary indication that the breakaway oxidation does not play role in design basis accident LOCA conditions with the E110 alloy. This is to be confirmed by transient oxidation tests.

SARNET benchmark on Phébus FPT3 integral experiment on core degradation and fission product behaviour

The importance of computer simulations in the assessment of nuclear plant safety systems has increased dramatically during the last three decades. The systems of interest include existing or proposed systems that operate, for example, normal operation, in design basis accident conditions, and in severe accident scenario beyond the design basis. The role of computer simulations is especially critical if one is interested in the reliability, robustness, or safety of high consequence systems that cannot be physically tested in a fully representative environment. In the European 7th Framework SARNET project, European Commission (EC) co-funded from 2008 to 2013, the Phébus FPT3 experiment was chosen as a code benchmark exercise to assess the status of the various codes used for severe accident analyses in light water reactors.The aim of the benchmark was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor (PWR), starting from the initial stages of core degradation, fission product, actinide and structural material release, their transport through the primary circuit up to the behaviour of the released fission products in the containment.The FPT3 benchmark was well supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production were well captured. No code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. The total volatile fission product release was well simulated, but the kinetics were generally overestimated. Concerning the modelling of semi-volatile, low-volatile and structural material release, the models need improvement, notably for Mo and Ru for which a substantial difference between bundle and fuel release was experimentally observed, due to retention in the cooler upper part of the bundle. The retention in the primary circuit was not well predicted, this was due mainly the non-prototypic formation of a boron-rich blockage in the rising line of the FPT3 steam generator, simulated in the circuit as a single external cooled U tube. The deposition mechanism and the volatility of some elements (Te, Cs, I) could be better predicted.Containment vessel thermal hydraulics, designed in the experiment to be well-mixed, were well calculated. Concerning the containment aerosol depletion rate, only stand-alone cases (in which the input data were derived from experimental data) provided acceptable results, whilst the integral cases (in which the input data came from circuit calculations) tended to largely overestimate the total aerosol airborne mass entering the containment.The disagreement of the calculated total aerosol airborne mass in the containment vessel with the measured one is due to the combination of a general underestimation of the overall circuit retention and overestimation of fission product and structural material release.Calculation of iodine chemistry in the containment turned out to be a major challenge. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit during the test, which is not correctly reproduced by the codes. This inability of the codes compromised simulation of the observed iodine behaviour in the containment.In the benchmark a significant user effect was detected (different results being obtained by different users of the same code) which had to be taken into account in analysing the results. This article reports the benchmark results comparing the main parameters calculated and observed, summarising the results achieved, and identifying the areas in which understanding needs to be improved. Relevant experimental and theoretical work is under way to resolve the issues raised.

Structural mechanical and thermal hydraulic aspects on the behaviour of crack like leaks in piping

The evaluation of fluid flow rates through crack-like leaks in pressurized components of NPPs plays an important role for leak-before-break considerations. When using available simplified methods for the estimation of critical discharge flow rates, a suitable friction factor has to be chosen. Guidelines are given in the German code KTA 3206 based on investigations of leak rate experiments summarized in this paper. The frictional pressure loss in different leak flow approaches has to be chosen consistently in order to obtain comparable results with different models.Thermal hydraulic and structural mechanical analyses were performed for a postulated leak in the pressurizer surge line (SL) of a PWR Konvoi type under design basis accident conditions. The leak was postulated in form of a through-wall crack in circumferential direction. The size of the leak was calculated in the framework of Finite Element calculations with the code ADINA using an analysis model of a cooling loop of Konvoi type. With the calculated leak size, an ATHLET calculation was conducted, also examining the influence of the consideration of a variable leak size depending on the system pressure and mechanical loads. In the investigated case, the implication of the decreasing leak size, especially on the course of the remaining system pressure, cannot be neglected. The reduction of the leak area amounts in the transient examined to ca. 25% after about 1 h due to both the system pressure and temperature decrease and leads therefore also to an approximately 23% smaller leak rate. These results are dependent on the assumed position of the leak.Concerning the determination of the leak rates as critical flow-through rates with simplified methods good agreement has been achieved although the approaches differ considerably in parts. Relevant influence factors are the inflow losses in the crack channel, the consideration of the hydraulic diameter in connection with the leak area and the assumptions on the flow resistance coefficient due to the roughness of the crack surfaces.

ICONE23-2171 Design Considerations for Insulation System for Fluoride Salt Cooled High Temperature Reactor under Beyond Design Basis Accident Conditions

ICONE23-2171 Design Considerations for Insulation System for Fluoride Salt Cooled High Temperature Reactor under Beyond Design Basis Accident Conditions

원자력발전소 직류전원계통용 축전지 성능시험 분석

Function of battery bank stores energy for DC load in general, and DC power system of the nuclear power plant is used to supply DC loads for safety- featured instrumentation and control such as inverter, class 1E power system control and indication, and station annunciation. Class 1E DC power system must provide a power for the design basis accident conditions, and adequate capacity must be available during loss of AC power and subsequent safe shutdown of the plant. In present, batteries of Class 1E DC power system of the nuclear power plant uses lead-acid batteries. Class 1E batteries of nuclear power plants in Korea are summarized in terms of specification, such as capacity, discharge rate, bank configuration and discharge end voltage, etc. This paper summarizes standards of determining battery size for the nuclear power plant, and analyzes duty cycle for the class 1E DC power system of nuclear power plant. Then, battery cell size is calculated as 2613Ah according to the standard. In addition, this paper analyzes performance test results during past 13 years and shows performance degradation in the battery bank. Performance tests in 2001 and 2005 represent that entire battery cells do not reach the discharge-end voltage. Howeyer, the discharge-end voltage is reached in 14.7% of channel A (17 EA), 13.8% of channel B (16 EA), 5.2% of channel C (6 EA) and 16.4% of channel D (19 EA) at 2011 performance test. Based on the performance test results analysis and size calculation, battery capacity and degradation by age in Korearn nuclear power plant is discussed and would be used for new design.

QUENCH-LOCA program at KIT on secondary hydriding and results of the commissioning bundle test QUENCH-L0

New out-of-pile QUENCH-LOCA bundle tests are being performed in the QUENCH facility within the Nuclear Safety Program of KIT. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design basis accident conditions as well as detailed post-test investigation of the mechanical properties of the claddings to check the embrittlement criteria. The program was started in 2010 with the QUENCH-L0 commissioning test using 21 electrically heated rods with as-received Zircaloy-4 claddings followed by the QUENCH-L1 reference test using the same material. Advanced cladding alloys (Duplex DX/D4, M5®, ZIRLO™) will be applied in further tests. For QUENCH-L0 each rod was separately pressurized with krypton with initial pressures from 35 to 55bar. The transient phase with heating from 793K to 1343K lasted 185s. The increased ductility of the heated cladding resulted in a progressive ballooning and consequent burst of all of the pressurized rods during the transient. The test was terminated by water quenching of the bundle without the usual intermediate slow cooling phase. Post-test investigations showed circumferential strain values between 20 and 35% at cladding positions with oxidation degree corresponding to 2% ECR. Neutron radiography of cladding tubes showed that at the inner rod simulator group so called secondary hydriding occurred. Maximal hydrogen concentrations up to 2500wppm on the boundary of the internal oxidized region around burst positions were determined. These hydrogen containing bands correlate with cladding bulk zones with increased micro hardness. XRD and TEM analysis indicated dissolution of hydrogen in the metal matrix without noticeable zirconium hydrides formation. Tensile tests of claddings with hydrogen containing bands showed simultaneous cladding tube ruptures at these bands below and above the burst opening.