Rig Of Safety Assessment Research Articles

Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test

In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR™, a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks.As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR™ and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR™ nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed very similar results as the experiment and the ROSA-TRACE calculation. The only significant difference observed was a faster primary depressurization after the break flow turned to single-vapor flow. This difference could be explained on the basis of geometrical differences between the EPR™ and ROSA/Westinghouse RPV's designs.

Comparative Study of Station Blackout Counterpart Tests in APEX and ROSA/AP600

A comparison is presented between station blackout tests conducted in both the Advanced Plant Experiment (APEX) facility and in the modified Rig of Safety Assessment (ROSA/AP600) Large-Scale Test Facility. The comparison includes the depressurization and liquid-level behavior during secondary-side blowdown, natural circulation, automatic depressurization system operation, and in-containment refueling water storage tank injection. Reasonable agreement between the test results from APEX NRC-2 and ROSA/AP600 AP-BO-01 has been observed with respect to the timing of depressurization and liquid draining rates. This indicates that the reduced height and pressure scaling of APEX preserves the sequence of events relative to the full-height and pressure ROSA/AP600.

Implications of the Rosa/AP600 High- and Intermediate-Pressure Test Results

Westinghouse has submitted the new AP600 reactor design for certification under the U.S. Nuclear Regulatory Commission (NRC) regulations for standard designs. The NRC has performed confirmatory testing in the Rig of Safety Assessment (ROSA) - V Large-Scale Test Facility run by the Japan Atomic Energy Research Institute (JAERI). The ROSA/AP600 test results as provided by JAERI are reviewed in terms of their implications for the original technical concerns that were to be addressed by the testing program. Implications for computer code capabilities are also discussed. Since gravity-driven natural circulation flow in a complicated piping network was a key concern, the review concentrates on this aspect of the ROSA/AP600 test results. In particular, it looks at the possible effect of system interactions at high pressure and during early depressurization. It identifies those ROSA/AP600 test occurrences that point to processes that could delay automatic depressurization system initiation or in-containment reactor water storage tank injection. Since most of the tests run were small-break loss-of-coolant accidents (SBLOCAs), the review focusses on this type of scenario. A comparison of several SBLOCA tests is presented.

RELAP5/MOD3 Simulation of the Loss of the Residual Heat Removal System During a Midloop Operation Experiment Conducted at the ROSA-IV Large-Scale Test Facility

A simulation of the loss of residual heat removal (RHR) system during midloop operations was performed using the RELAP5/MOD3 thermal-hydraulic code. The experiment was conducted at the Rig of Safety Assessment (ROSA)-IV/Large-Scale Test Facility. The experiment involved a 5% cold-leg break along with the loss of the RHR system. The transient was simulated for 3040 s. Core boiling and subsequent primary system pressurization occurred after the initiation of the transient. There was a good agreement between the measured and the calculated data until the loop seal clearing (LSC). It was found that the steam condensation was underpredicted in the calculations. This caused the calculated data after the LSC to differ from that of the measured data. The core rod surface temperature excursion around the occurrence of the LSC was not calculated. Overall, there was good qualitative agreement between the measured and the calculated data. The calculations, performed on the CRAY-YMP supercomputer, took over 60 h of CPU time for a transient of 51 min.

BWR Loss-of-Coolant Accident Tests at ROSA-III with High Temperature Emergency Core Coolant Injection

The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-III integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests.

BWR natural circulation experiments at ROSA-III under low-power and reduced inventory conditions.

The natural circulation core cooling is important during small break loss-of-coolant accidents and transients involving loss of forced-circulation core cooling. This paper presents results of natural circulation experiments conducted in the ROSA-III (Rig of Safety Assessment No. 3) facility under reduced mass inventory conditions. Three distinct modes of BWR natural circulation were identified in the experiments: the primary circulation mode, the internal circulation mode and the open-loop (core uncovery) mode. One-dimensional quasi-steady analytical model was developed for the internal natural circulation mode to determine the core uncovery criterion. The model predicted well the two-phase mixture level inside the core shroud for quasi-steady state natural circulation at pressures of 7.35 and 2.06 MPa, and scaled core powers below 20%, and was used to predict BWR natural circulation behaviors.

SAFER03 and TRAC-BD1 Analyses of a ROSA-III Large-Break Experiment on a Boiling Water Reactor

Simulation tests of a boiling water reactor (BWR) loss-of-coolant accident (LOCA) caused by a double-ended guillotine break of a recirculation pump suction line were carried out with the rig of safety assessment III (ROSA-III) test facility at the Japan Atomic Energy Research Institute. SAFER03 and TRAC-BD1 analyses with the ROSA-III test data have been performed to assess predictive capability for the large-break LOCA thermal-hydraulic response of a BWR. The analytical results indicate that SAFER03 and TRAC-BD1 predicted key phenomena very well. Furthermore, it was confirmed that SAFER03 predicted higher peak cladding surface temperature than the test data due to the conservative thermal-hydraulic model.

Core liquid level depression due to manometric effect during PWR small break LOCA. Effect of break area.

The 10, 5 and 2.5% cold leg break loss-of-coolant accident experiments were conducted by using the Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program. In the early stage of the 5% break experiment, the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed. However, the core liquid level depression without the core dry out was observed in the 10 and 2.5% break experiments. In the three break experiments, the core liquid levels were recovered just after the loop seal clearing. The manometric effect due to the liquid seal formation in the loop seal and the liquid holdup in the steam generator (SG) U-tubes upflow-side caused a depression of the core collapsed liquid level. The liquid holdup in the U-tubes upflow-side was observed after the termination of the two-phase circulation due to the phase separation at the U-tubes top. The counter current flow limiting (CCFL) and the condensation of steam was considered to be the mai...

Effect of heat generation difference among fuel bundles on core thermal-hydraulics during 200% and 5% loss-of-coolant accident experiments at ROSA-III.

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6)system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS)tests. Two 200% and two 5% recirculation pump suction line break LOCA experiments with and without heat generation difference among the bundles were conducted in order to investigate interaction of thermal-hydraulic phenomena in a multi-bundle core. The axial propagation of quenching tends to become more nonuniform in the same bundle as well as in the core in both large and small break experiments when the heat generation difference among the bundles exists. The small flow imbalance among the bundles which is caused by void fraction difference due to the heat generation difference among the bundles is estimated to make the flow condition nonuniform and more agitated and increase the radial scattering of the quenching propagation.

Effect of Heat Generation Difference among Fuel Bundles on Core Thermal-Hydraulics during 200% and 5% Loss-of-Coolant Accident Experiments at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6)system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS)tests. Two 200% and two 5% recirculation pump suction line break LOCA experiments with and without heat generation difference among the bundles were conducted in order to investigate interaction of thermal-hydraulic phenomena in a multi-bundle core. The axial propagation of quenching tends to become more nonuniform in the same bundle as well as in the core in both large and small break experiments when the heat generation difference among the bundles exists. The small flow imbalance among the bundles which is caused by void fraction difference due to the heat generation difference among the bundles is estimated to make the flow condition nonuniform and more agitated and increase the radial scattering of the quenching propagation.

Investigation of effect of pressure control system on BWR LOCA phenomena using ROSA-III test facility.

A pressure control system failure test series was conducted at the Rig of Safety Assessment (ROSA)-III test facility to evaluate the effect of the pressure control system on thermal-hydraulic phenomena during a small break loss-of-coolant accident (LOCA) of a boiling water reactor (BWR). The break was assumed at the recirculation pump suction line. The pressure control system had no effect for breaks greater than 5%. For breaks less than 5%, if the pressure control system was inactive, the core was uncovered temporarily because of the bubble collapse due to pressure rise by the closure of the main steam isolation valve (MSIV), and the fuel rod surface temperature rose high during this period. However, the peak cladding temperature (PCT), which occurred mainly during the later core uncovering by boil-off, was lower in a LOCA with pressure control system failure than in a corresponding LOCA with intact pressure control system. This is because the emergency core cooling systems (ECCSs) were actuated earlier in a LOCA with pressure control system failure due to lower system pressure. The PCT was well below the present safety criteria of 1,473K even if the pressure control system and high pressure core spray system (HPCS) (the severest single failure in ECCS) were assumed to be inactive.

ＲＯＳＡ‐ＩＶ計画の大型非定常試験装置（ＬＳＴＦ）における実験の開始

The Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV Program is an integral test facility to investigate thermal-hydraulic response of a pressurized water reactor (PWR) system during small break loss-of-coolant accidents (LOCAs) and operational transients. The LSTF is volumetrically scaled at 1/48 and has the same height as the PWR. The construction of the LSTF was completed in May 1985 and several small break LOCAs, TMI-2 accident type simulation and natural circulation tests have been successfully conducted. Presented in the present report are outline of the test program at the LSTF and some test results including transient core liquid level depression during cold leg small-break LOCAs, break orientation effect in cold leg small-break LOCAs on the system transient and thermal-hydraulic phenomena during the TMI-2 type multifailure accident simulation test.

Similarity study of large steam line break LOCA's in ROSA-III, FIST and BWR/6

Similarity of the thermal hydraulic phenomena in a 100% steam line break loss-of-coolant accident (LOCA) between the Rig-of-Safety Assessment (ROSA)-III. Full-Integral Simulation Test (FIST) and a boiling water reactor (BWR)/6 system has been studied experimentally and analytically. The experimental results of ROSA-III (RUN952) and FIST (6MSB1) showed similar LOCA phenomena except for the core cooling. The core cooling was affected by the different ECCS actuation logics used in the tests. The effects of the different test conditions and the system-inherent features on the LOCA phenomena were separately evaluated through the post-test and similarity analysis of the ROSA-III and FIST tests by using RELAP5/MOD1 code with a jet pump model. The similarity of the major events in the ROSA-III and FIST facilities to those of BWR/6 system were confirmed assuming the same ECCS actuation logic and the same sealed initial mass inventory among the three systems. Differences in vessel geometries, metal stored heat and core power curves caused slight differences in the responses of pressure and fuel surface temperatures.

Blind-Blind Prediction by RELAP5/MOD1 for a 0.1% Very Small Cold-Leg Break Experiment at ROSA-IV Large-Scale Test Facility

The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for themore » analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily.« less

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III.

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.

Recirculation Pump Discharge Line Break Tests at ROSA-III for a Boiling Water Reactor

Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (A /SUB j/ + A /SUB p/ ) for discharge line breaks and (A /SUB j/ + A /SUB o/ ) for suction line breaks, where A /SUB j/ , A /SUB p/ , and A /SUB o/ are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the more » fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power. « less

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Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.

Experimental analysis of the power curve sensitivity test series at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used. In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment.

ROSA-III Double-Ended Break Test Series for a Loss-of-Coolant Accident in a Boiling Water Reactor

The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.