Experimental and calculation results of the integral reflood test QUENCH-15 with ZIRLO™ cladding tubes in comparison with results of previous QUENCH tests

The QUENCH-15 experiment investigated the effect of ZIRLO™¹ cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (standard Zircaloy-4), QUENCH-12 (VVER, E110), and QUENCH-14 (M5^®). The QUENCH-15 bundle cross-section corresponded to a Westinghouse PWR core design and consisted of 24 heated rods (internal tungsten heaters between 0 and 1024 mm axial elevation, cladding oxidised region between −470 and 1500 mm), six corner rods made of Zircaloy-4, two corner rods made of E110, and a Zirconium 702 shroud. The test was conducted in principle with the same protocol as QUENCH-06, -12 and -14, so that the effects of the change of cladding material and bundle geometry could be more easily observed. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1473 K over a period of about 3000s. The power was then ramped at a rate of 0.25 W/s/rod to cause a temperature increase until the desired maximum bundle temperature of 2153 K was reached. The maximum oxide layer thickness observed was 380 μm. Then reflood with 1.3 g/s/rod water at room temperature was initiated. The electrical power was reduced to 175 W/rod during the reflood phase, approximating effective decay heat level. The post-test metallography of the bundle showed neither noticeable breakaway oxidation of the cladding nor melt release into space between rods. The average outer oxide layer thickness at hottest elevation of 950 mm was 620 μm (QUENCH-06: 630 μm). The molten cladding metal at hottest elevation was localised between the outer and inner oxide layers. The thickness of inner oxide layer reaches 20% of that of the outer oxide layer. The measured hydrogen release during the QUENCH-15 test was 41 g in the pre-oxidation and transient phases and 7 g in the quench phase which are comparable with those in QUENCH-06, i.e. 32 g and 4 g, respectively. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2. The calculation results support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5^® or ZIRLO™ for the beyond-design basis accident present conditions here studied.     

Experimental and calculation results of the integral reflood test QUENCH-14 with M5® cladding tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5® cladding material on hydrogen generation, in comparison with Zircaloy-4.     

AgInCd control rod failure in the QUENCH-13 bundle test

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO₂ pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary.Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H₂ during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m).Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of 2300 K (compared to 1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.     

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA.     

Reflooding experiments with LWR-type fuel rod simulators in the QUENCH facility

The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO₂. After a transient phase with a heating rate of 0.5–1 K s⁻¹ water of approx. 395 K is admitted from the bottom when the test bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s⁻¹ each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 μm oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre-oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of 1870 K. In the second bundle experiment, QUENCH-02, quenching started at 2500 K. Pre-oxidation apparently prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behavior of the two experiments is also reflected in hydrogen generation. While the bulk of H₂ was produced during pre-oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH-02, at a maximum production rate of 2.5 g s⁻¹ as compared to 0.08 g s⁻¹ in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behavior, the quench temperatures, and quench velocities.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

Analysis of International Standard Problem No. 45, QUENCH06 Test at FZK by Detailed Severe Accidents Analysis Code,IMPACT/SAMPSON

The QUENCH06 test at Forschungszentrum Karlsruhe (FZK) selected to be the international standard problem (ISP) No. 45 by OECD/NEA has been analyzed by the IMPACT/SAMPSON code, a detailed analysis code for severe accidents in an LWR. The code has been used to perform blind and open phase analyses of the quench phenomenon and Zr/water reaction on the actual PWR cladding. Conclusions obtained from the analyses are as follows: (1) Overall, the blind phase analysis predicted relatively well the results of QUENCH06 test with respect to thermal hydraulics in the test bundle and its degradation due to Zr/water reaction. (2) The difference in the final accumulated hydrogen generation between the blind phase analysis and the test was 19% under parabolic rate constants by Cathcart-Pawel and Urbanic-Heidrick with a transition temperature of 1,853 K. (3) Embrittlement failure criteria for Zircaloy cladding against quenching adopted from the SCDAP/RELAP5 (MATPRO) predicted almost correctly the failure time of the QUENCH06 test. (4) The open phase analysis has traced the temperature change of the inner fuel cladding very well and has yielded almost the measured hydrogen release, considering inner surface oxidation after cladding failure. (5) The IMPACT/SAMPSON code has been validated against the quench phenomenon and the Zr/water reaction observed in the QUENCH06 test at FZK.     

Assessment of RELAP5/MOD3.2.2 Gamma against ABB Atom 3×3-Rod Bundle Reflooding Tests

The results of the ABB Atom 3×3-Rod Bundle Reflooding Tests were used for assessment of the reflooding model used in RELAP5/MOD3.2.2 Gamma version. The assessment calculations were performed using the default calculation model options implemented in the code.The tests were performed to investigate the effects of different spacer grid designs on heat transfer during the reflooding period of a pressurized water reactor loss-of-coolant accident (LOCA). The tests were conducted under low-pressure and low-flow (LPLF) conditions using a PWR-type 3×3-rod bundle with full-length indirectly electrically heated, stepped cosine axial power-shaped heater rods. Three different spacer grid configurations were studied: spacer grids without mixing vanes, mixing vane spacer grids, and mixing vane spacer grids together with intermediate flow mixers (IFM).A total of 36 tests with different spacer grid configurations were calculated. For two selected basic tests with non-mixing spacer grids an extended comparison of calculated and measured parameters is presented. The comparison of the predicted and measured maximal cladding temperatures and quench times, which are the most important parameters in licensing calculations, is presented for all the performed tests.The assessment calculations were preceded by nodalization, time step, and moving mesh studies.The RELAP5/MOD3.2.2 Gamma code was found to still have several deficiencies in the reflood model. The calculation results show a satisfactory agreement with experimental inner peak cladding temperature, however the predicted temperature turn-around times and quench times are significantly too short. The results also show a significant over-prediction of the reflood heat transfer and the vapour temperatures. The void profile downstream the quench front is not correctly predicted either. Finally, the present reflood model does not properly reflect the effects of spacer grids on the reflood heat transfer.In spite of these deficiencies the improvements incorporated into RELAP5/MOD3.2 by the Paul Scherrer Institute (PSI) eliminated the unphysical behaviors such as continuous cooling without clear turn-around temperature and no visible quenching phenomena, which were shown in the reflood calculations by means of the RELAP5/MOD3.1 code.     

Separate-effect tests on zirconium cladding degradation in air ingress situations

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident.A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available.New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program—ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the “as-received” state, or after pre-oxidation in steam. “Analytical” tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 °C up to 1500 °C. Systematic post-test metallographic inspections are performed.The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale degradation under air exposure, important features that have to be taken into account are highlighted.     

Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues.A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process.A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed to study fuel bundle exposure to steam first and then to air. This paper deals with the main results obtained with MAAP4.07 when simulating QUENCH-10.     

Advanced treatment of zircaloy cladding high-temperature oxidation in severe accident code calculations: Part III. Verification against representative transient tests

The treatment of zirconium oxidation kinetics in severe accident (SA) codes has been the subject of many discussions and controversies in recent years. The main problem was the existence of several correlations which could lead to large differences in the calculated results. It appeared clearly that there was a need to converge towards a common understanding of the physical processes that must be modeled (oxygen diffusion, blanketing effect, etc.) and an agreed database among code developers and users. It would help reducing an important source of uncertainties in SA calculations.The kinetic correlation database, obtained as a result of examination of complementary experimental data in Parts I and II, is applied here to analyze a few high-temperature separate-effects tests and bundle experiments where Zry oxidation reaction played a dominant role. The ICARE/CATHARE computer code developed by IRSN is used to check the validity of the high-temperature correlations derived in Parts I and II. The physical modeling provided by the code includes detailed account of specific features of chemical interactions between fuel rod cladding and steam. In particular, high reaction rates at T > 2000 K are moderated by two effects, examined in Part I: steam blanketing during thin oxide layers growth and transition to oxidation of α-Zr(O) phase after total consumption of primary β-Zr in cladding metallic part.When applied to separate-effects tests, the evaluated parabolic correlations have shown their applicability to different types of temperature transients taking into account Zry oxidation specifics in rod geometry. The bundle integral experiments QUENCH-06 and PHEBUS B9+ did not lead to extremely large temperature excursions. Calculated temperatures, hydrogen production and oxide thickness, as well as parameters of melt relocation were found to agree well with experimentally measured values. As a result of this study, we believe that the new best-fitted correlations, obtained in agreement with available experimental data, can be used in further studies and can improve predictive power of the codes. The continuation of the current work will be the application of ICARE/CATHARE code with best-fitted Zry oxidation correlations to NPP accident scenarios.     

Void Fraction in Simulated PWR Fuel Bundle during Reflood Phase

In order to evaluate void fraction in a bundle geometry during the reflood phase, reflooding experiment with a 4×4 simulated fuel array was conducted.

As the result, it was found that the effects of the clad temperature and the power of the heater rods are small and the effects of the pressure and the inlet flow rate are large on the relationship between the superficial steam velocity and the void fraction in a bundle geometry during the reflood phase. It was, also, found that there is no distinct difference of the void fractions caused by the different flow patterns in the wet clad region and in the dry clad region in a bundle geometry during the reflood phase, when compared at the same superficial steam velocities.

Furthermore, the applicability of Cunningham-Yeh's void fraction correlation was investigated under a wide range of conditions anticipated during the reflood phase. The range of conditions under which Cunningham-Yeh's correlation predicts the void fraction within an error band of ±20% were made clear.     

环形燃料棒束再淹没行为实验研究

环形燃料是一种可在维持或提升安全裕度的前提下大幅提高反应堆经济效益的新型压水堆燃料,由于其双面冷却的特点,环形燃料在LOCA再淹没阶段的热工水力行为与传统实心燃料存在显著差异。现有关于环形燃料再淹没行为的实验研究鲜有报道。本研究基于自主设计的高温环形电加热棒建立了环形棒束再淹没实验装置,开展了3×3环形棒束底部再淹没实验研究,探究了环形棒束再淹没典型物理过程及不同工况下再淹没关键参数的变化规律。结果表明,环形棒束再淹没物理过程与传统实心棒束类似,且内外通道的骤冷前沿推进和传热模式变化趋于同步。在同一时刻下,环形棒内外壁面间存在温度梯度。骤冷前沿推进速度随再淹没速度和过冷度的增大而增大,随峰值包壳温度和线功率密度的增大而减小。此外,定位格架在低流速、低过冷度与高壁温工况下能显著提升下游的骤冷前沿推进速度。     

High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

Oxidation experiments were conducted at 1000-1200 °C in flowing steam with samples of as-received Zr-1Nb alloy E110 tubing and/or polished E110 tubing. The purpose was to determine the oxidation behavior of this alloy under postulated loss-of-coolant accident conditions in light water reactors. The as-received E110 tubing exhibited a high degree of susceptibility to nodular oxidation and breakaway oxidation at relatively low test times, as compared to other cladding alloys. The nodules grew much more rapidly at 1000 °C than 1100 °C, as did the associated hydrogen uptake. The oxidation behavior was strongly affected by the surface condition of the materials. Polishing to ≈0.1 μm roughness (the roughness of the as-received tubing was ≈0.4 μm) delayed breakaway oxidation. Polishing also removed surface impurities. For polished samples oxidized at 1100 °C, no significant nodular oxidation appeared up to 1000 s. For polished samples oxidized at 1000 °C, hydrogen uptake >100 wppm was delayed from ≈300 s to >900 s. Weight-gain coefficients were determined for pre-breakaway oxidation of polished-only and machined-and-polished E110 tubing samples: 0.162 (mg/cm²)/s^0.5 at 1000 °C and 0.613 (mg/cm²)/s^0.5 at 1100 °C.     

Quenching of high temperature VVER fuel after long term oxidation in hydrogen rich steam

A cleaning tank incident at Paks NPP resulted in severe fuel damage of thirty VVER-440 type assemblies. The fuel rods heated up due to insufficient cooling and the zirconium components suffered heavy oxidation. Opening of the tank and quenching of the assemblies by cold water led to fragmentation of brittle zirconium components. In order to improve the understanding of the phenomena that took place during the Paks-2 incident, integral tests with electrically heated fuel rods have been carried out. The experiments covered the whole scenario of the incident, including the long term oxidation in hydrogen rich steam atmosphere and final quenching by cold water. The final state of the bundles was very brittle, the fuel rods were fragmented and they showed many similarities with the results of the incident at the NPP. For this reason, it is very probable that the thermal conditions and chemical reactions were also similar in the tests and in the incident. The maximum temperature in the cleaning tank was probably in the range of 1200–1300 °C. The hydrogen in the Zr components could reach 20,000 ppm, while the hydrogen concentration in the atmosphere of the cleaning tank could be above 80 vol.%. The post-test examination of test bundles indicated that the high degree of embrittlement was a combined result of oxidation and hydrogen uptake by the Zr components.     

Effect of SAMG entry condition on operator action time for severe accident mitigation

The effect of SAMG (Severe Accident Management Guidance) entry condition on operator action time for prevention of reactor vessel failure in the OPR1000 (Optimized Power Reactor 1000) was evaluated with a comparison to other conditions using the SCDAP/RELAP5/MOD3 computer code in detail. Dominant severe accident sequences, such as a station blackout, a total loss of feedwater, a small and medium break loss of coolant accidents without a safety injection were evaluated in this study. The results showed that a core exit temperature of 480 °C for the SAMG entry condition was too early, but a core exit temperature of 1000 °C was too late for operator action time to prevent reactor vessel failure. The SAMG entry condition of a core exit temperature of 650 °C for the OPR1000 was suitable for the operator action timing point in order to mitigate a severe accident.     

Evaluation of the RCS depressurization strategy for the high pressure sequences by using SCDAP/RELAP5

As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough.     

Degradation and oxidation of B4C control rod segments at high temperatures

Extensive series of test were performed of the degradation of boron carbide absorber rods and the oxidation of the resultant absorber melts. Various types of control rod segments made of commercial materials used in French 1300 MW PWRs were investigated in the temperature range between 800 °C and 1700 °C in a steam atmosphere. The gaseous reaction products were analyzed quantitatively by mass spectroscopy for evaluation of the oxidation rates. Extensive post-test examinations were performed by light microscopy, scanning electron microscopy as well as EDX and Auger spectroscopy. Rapid melt formation due to eutectic interactions of stainless steel (cladding tube) and B₄C, on the one hand, and steel and Zircaloy-4 (guide tube), on the other hand, was observed at temperatures above 1250 °C. Complex multi-component, multi-phase melts were produced. ZrO₂ oxide scale on the outside kept the melt within the guide tube, thus preventing its early relocation and oxidation. Rapid oxidation of the absorber melts and remaining boron carbide pellets took place after failure of the protective oxide shell above 1450 °C. Only very little methane was produced in these tests which is of interest in fission product gas chemistry because of the production of organic iodine.     

Experiments on air ingress during severe accidents in LWRS

Various kinds of experiments on the oxidation of Zircaloy-4 cladding material in different scales and under different conditions at temperatures 800–1300 °C (small scale) and up to 2000 °C (large scale) are presented. The focus of this work was on prototypic mixed air–steam atmospheres and sequential reaction in steam and air, where no data were available before. The separate-effects tests were performed to support the large scale bundle test QUENCH-10 and to deliver first data for model development.     

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.